ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Nevada Test Site

– Occupational Environmental Dose

Document Number:

ORAUT-TKBS-0008-4 

Effective Date: 12/08/2006

Revision No.: 00 PC-1 

Controlled Copy No.: ______

Page 1 of 51

Subject Expert: Eugene Rollins

Document Owner

Approval: Signature on File Date: 04/15/2004

Eugene Rollins, TBD Team Leader

Approval: Signature on File Date: 04/14/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 04/19/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 04/20/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

Acronyms and Abbreviations ...............................................................................................................

Page

4

4.1 Introduction................................................................................................................................. 6

4.1.1 Purpose..............................................................................................................................7

4.1.2 Scope.................................................................................................................................7

4.2 Internal Dose from OnSite Atmospheric Radionuclide Concentrations ........................................ 7

4.2.1 On-Site Releases to Air......................................................................................................7

4.2.1.1 Source Description ....................................................................................................7

4.2.1.1.1 Weapons Testing ..............................................................................................7

4.2.1.1.2 Ground Seepage of Noble Gases ..................................................................... 8

4.2.1.1.3 Tunnel Operations ............................................................................................9

4.2.1.1.4 Containment Ponds ..........................................................................................9

4.2.1.1.5 Drillbacks ..........................................................................................................9

4.2.1.1.6 Laboratories......................................................................................................9

4.2.1.1.7 Radioactive Waste Management Sites.............................................................. 9

4.2.1.2 Atmospheric Radionuclide Concentrations .............................................................. 10

4.2.1.2.1 Tritium............................................................................................................. 11

4.2.1.2.2 Plutonium........................................................................................................ 14

4.2.1.2.3 Gross Alpha .................................................................................................... 21

4.2.1.2.4 Gross Beta...................................................................................................... 24

4.2.2 Annual Intakes from Resuspension .................................................................................. 24

4.2.3 Uncertainty.......................................................................................................................32

4.2.3.1 Time-Dependent Nature of the Resuspension Factor .............................................. 32

4.2.3.2 Spatial Variations in Radionuclide Soil Concentrations ............................................ 32

4.2.3.3 Duration of Exposures ............................................................................................. 33

4.3 Occupational External Dose...................................................................................................... 33
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4.3.1 Ambient Radiation............................................................................................................ 33

4.3.2 Releases of Noble Gases................................................................................................. 38

4.4 Occupational Dose to Tunnel Workers from Exposure to Radon............................................... 39

4.4.1 Tunnel Activities ............................................................................................................... 40

4.4.2 Radon Measurements ...................................................................................................... 40

4.4.3 Tunnel Radon Concentrations.......................................................................................... 41

4.4.4 Tunnel Workers Exposure to Radon................................................................................. 43

4.4.5 Uncertainty.......................................................................................................................43

4.5 Instructions to the Dose Reconstructor ..................................................................................... 44

4.5.1 Assigning External Ambient Dose .................................................................................... 44

References ........................................................................................................................................46

LIST OF TABLES

Table Page

4.2.1.2.1-1 Historical HTO atmospheric concentrations by year and Area with 

estimated maximum and average annual organ dose................................................. 13

4.2.1.2.2-1 239Pu atmospheric concentrations for sampled Areas ................................................. 15

4.2.1.2.2-2 Annual 239Pu intakes for sampled Areas ..................................................................... 17

4.2.1.2.2-3 238Pu atmospheric concentrations for sampled Areas ................................................. 19

4.2.1.2.2-4 238Pu intakes for sampled Areas ................................................................................. 20

4.2.1.2.3-1 Gross alpha atmospheric concentrations for sampled Areas ...................................... 22

4.2.1.2.3-2 Gross alpha annual intakes for sampled Areas........................................................... 23

4.2.1.2.4-1 Gross beta atmospheric concentrations for sampled Areas ........................................ 25

4.2.1.2.4-2 Gross beta annual intakes for sampled Areas ............................................................ 27

4.2.2-1 Radionuclide areal soil deposition .............................................................................. 30

4.2.2-2 Average annual intakes for NTS Areas from soil resuspension................................... 31

4.2.2-3 Site average and maximum annual intakes for NTS from soil 

resuspension..............................................................................................................31

4.3.1-1 Ambient background radiations levels by year............................................................ 35

4.3.1-2 Ambient radiation by Area .......................................................................................... 36

4.3.1-3 85Kr average background and the Area 20 net concentrations and 

annual organ doses from 1983 through 1992 ............................................................. 39

4.4.3-1 Results of experiment to determine the effects of ventilation 

conditions on RDCs in the G Tunnel inclined drift....................................................... 41

4.4.3-2 Radon daughter concentrations for N-Tunnel in 1991 and 1992................................. 42

4.4.3-3 Radon daughter concentrations for P-Tunnel in 1991 and 1992 ................................. 42

4.4.4-1 Annual 222Rn exposures and uncertainties for internal dose 

reconstruction for miners and tunnel workers ............................................................. 43

LIST OF FIGURES

Figure Page

4.2.1.2-1 Air sampling stations in 1997 12

4.2.2-1 Resuspension factor as a function of time after initial deposition 30
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

04/20/2004 04/20/2004 00 New technical basis document for the Nevada 

Test Site– Occupational Environmental Dose. 

First approved issue. Initiated by Eugene M. 

Rollins.

04/20/2004 12/08/2006 00 PC-1 Approved page change revision as a result of 

biennial review. Adds acronyms and 

abbreviations on pages 4 and 5. Updates 

required language on page 6 in the Introduction. 

Adds Purpose and Scope sections on page 7. 

Added instructions to dose reconstructors on 

pages 44 and 45 in Section 4.5. As a result of 

formal internal review, deletes Section 4.5.2 from 

page 45. Completes references on page 47. 

This revision results in no change to the 

assigned dose and no PER is required. Training 

required: As determined by the Task Manager. 

Initiated by Eugene M. Rollins. Approval:

Signature on File________________ Eugene M. Rollins, Document Owner12/05/2006

Signature on File________________ John M. Byrne, Task 3 Manager10/24/2006

Signature on File _______________ Edward F. Maher, Task 5 Manager10/24/2006

Signature on File________________ Kate Kimpan, Project Director11/15/2006

Brant A. Ulsh Signature on File for__ James W. Neton, Associate Director for Science12/08/2006
Effective Date: 12/08/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-4 Page 4 of 51

ACRONYMS AND ABBREVIATIONS

Bq becquerel

CEDE committed effective dose equivalent

Ci curie

cm centimeter

d day

DoD U.S. Department of Defense

DOE U.S. Department of Energy

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPA U.S. Environmental Protection Agency

ft foot

GSD geometric standard deviation

hr hour

HT airborne tritium

HTO tritium oxide

in. inch

kg kilogram

km kilometer

L liter

LANL Los Alamos National Laboratory

lb pound

m meter

Mev megavolt-electron, 1 million electron volts

mg milligram

mi mile

min minute

ml milliliter

mR milliroentgen

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NTS Nevada Test Site

pCi picocurie

PERM passive environmental radon monitor

POC probability of causation

REECo Reynolds Electrical & Engineering Co., Inc.

RDC radon daughter concentration 

RPISU radon progeny integrating sampling unit
Effective Date: 12/08/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-4 Page 5 of 51

s second

Sv sievert

TLD thermoluminescent dosimeter

U.S.C. United States Code

WL working level

WLM working level month

yr year

μCi microcurie

§ section or sections
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4.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions for particular sites or categories of sites. The documents will be revised in the 

event additional relevant information is obtained about the affected site(s). These documents may be 

used to assist NIOSH staff in the completion of the individual work required for each dose 

reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
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• Radiation from diagnostic X-rays received in the treatment of work-related injuries

4.1.1

This TBD discusses occupational environmental dose. Occupational environmental dose is the dose 

received by individuals at the Nevada Test Site (NTS or Test Site) while outside onsite operational 

facilities during work activities. These doses can be internal or external depending on the 

characteristics of the individual radionuclides. While inhalation of most radionuclides would cause a 

dose to various organs in the body, noble gases would cause only an external dose because these 

inert radionuclides are not readily absorbed by the body. Job classifications of covered employees 

that may spend a majority of their time in the outdoor environment include carpenters, cement 

masons, drillers, electrical wiremen/linemen, ironworkers, laborers, operating engineers, painters, 

plumbers, pipefitters, sheet metal workers, and teamsters.

Purpose

4.1.2

Section 4.2 discusses internal doses to unmonitored workers from on-site releases to the air and 

resuspension of radioactive materials in soil. Section 4.3 describes external doses to workers from 

ambient radiation and releases of radioactive noble gases into air. Section 4.4 discusses dose to 

tunnel workers from exposure to radon.

Scope

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

Section 4.2.1 discusses the internal dose for workers outside the facilities, as determined from air 

concentrations resulting from ground-level releases. Section 4.2.2 discusses the internal dose from 

resuspension of radioactive materials in the soil. Section 4.2.3 discusses the uncertainties involved in 

these dose estimates.

4.2.1

4.2.1.1 Source Description

On-Site Releases to Air

4.2.1.1.1 Weapons Testing

NTS has been the primary location for the testing of nuclear explosives in the continental United 

States since 1951. Test programs have included atmospheric testing in the 1950s and early 1960s, 

earth-cratering experiments, and open-air nuclear reactor and rocket engine testing. Since the 

mid-1960s, testing of nuclear devices has occurred underground in drilled vertical holes or in mined 

tunnels. No nuclear tests have been conducted since September 1992. 

In all, more than 900 nuclear tests have taken place at the Test Site as part of these programs. One 

result of these tests is that the surface soils in many parts of NTS contain measurable amounts of 

several long-lived radionuclides. Almost all of the more than 100 above-ground tests contaminated 

the soil near ground zero. In addition, several underground tests were cratering experiments that 

threw radioactive rock and soil hundreds of feet, and some deeper underground tests vented 

radioactive material to the surface. A few safety tests, in which a nuclear device was destroyed by 

conventional explosives, scattered plutonium (and in some cases uranium) over the nearby ground. 

Finally, there was fallout of radioactive debris from many tests over the northern and eastern parts of 

the Test Site.
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Radiation levels at NTS have been monitored regularly, and safety officials have identified and fenced 

off areas where the soil is heavily contaminated. In many other areas, radionuclide levels are not high 

enough to warrant closing the area, but they are still above background. 

Atmospheric weapon and safety tests from 1951 and 1963 resulted in the release of approximately 

1.2 × 1010 Ci to the atmosphere. Much of this activity was from relatively short-lived radionuclides that 

decayed in a matter of days or weeks. The volatile radionuclides (such as radioiodines, noble gases, 

and tritium) were diluted in the atmosphere and transported off the site. However, much of the 

nonvolatile, long-lived radionuclides settled back into the soils at various locations at NTS (see 

Table 4.2.2-1, Section 4.2.2 below). These contaminated soils continue to represent a potential 

inhalation pathway to workers from resuspension of soils by wind and such mechanical activities as 

cleanup and remediation.

In 1963, nuclear weapons testing was moved underground to prevent the release of radionuclides to 

the atmosphere to achieve containment. Releases of radioactive material following an underground 

test are generally categorized with terms that describe both the volume of material released and the 

conditions of the release:

• Containment Failures: Containment failures are unintentional releases of radioactive matter to 

the atmosphere due to failure of the containment system. A prompt massive release, or one 

that occurs soon after a test, is a venting. 

• Late-Time Seeps: Late-time seeps are small releases that occur days or weeks after a test 

when gases diffuse through pore spaces in the overlying rock and are drawn to the surface by 

decreases in atmospheric pressure.

• Controlled Tunnel Purging: A controlled tunnel purge is an intentional release to allow either 

recovery of experimental data and equipment or reuse of part of the tunnel system.

• Operational Release: Operational releases are small consequential releases that occur when 

core or gas samples are collected, or when a drillback hole is sealed.

At the present time, processing of radioactive materials at NTS includes only laboratory analyses. 

Handling of these materials includes only the transport and storage of nuclear explosive devices and 

the operation of a radioactive waste management site for low-level and mixed wastes. Monitoring and 

evaluation of the various activities indicate that the potential sources of onsite radiation exposure are

releases from the following sources: 

• Evaporated tritiated water (HTO) from drainage containment ponds for E Tunnel in Area 12

• Onsite radioanalytical laboratories

• Area 3 and 5 waste facilities

• Other diffuse sources

The following sections describe effluent sources at NTS.

4.2.1.1.2 Ground Seepage of Noble Gases

Ground seepage can be enhanced when changes in ambient pressure pump small amounts of noble 

gases up through the overburden and into the atmosphere from the cavity created by a nuclear test. 

This process, sometimes referred to as atmospheric pumping, creates a diffuse source of radiological 

effluents. These area sources are rare and therefore not routinely monitored. The phenomenon is 
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usually restricted to tests conducted in the Pahute Mesa region of NTS. These seepages are from 

nuclear tests before 1993. 

4.2.1.1.3 Tunnel Operations

Nuclear tests occurred in mined tunnel complexes in the Rainier Mesa region. Because some tunnels 

were sealed in the mid 1990s, currently small amounts of contaminated water continue to drain from 

only one tunnel (Section 4.2.1.1.4). 

4.2.1.1.4 Containment Ponds

Water contaminated with radionuclides seeps from the tunnels in Area 12 and collects in containment 

ponds where some evaporates and some seeps into the soil. The only radiological contaminant 

which produces a measurable air emission from evaporation of the water is tritium (3

H) in the form of 

HTO. 

4.2.1.1.5 Drillbacks

Following underground nuclear tests, core samples have been taken from the cavity formed by the 

nuclear detonation for analysis and diagnosis. This core sampling is accomplished by drilling into the 

area of interest and recovering samples using special equipment. Radioactive material can escape to 

the atmosphere during these operations. 

4.2.1.1.6 Laboratories

The Reynolds Electrical & Engineering Company (REECo) conducted radiological analyses in 

Building 650, and Los Alamos National Laboratory (LANL) conducted similar analyses in Building 701 

at Mercury. Because these facilities have processed primarily environmental samples, very little 

radioactivity has passed through them. However, there was potential for some radionuclides to be 

discharged to the atmosphere through the hood ventilation system during sample processing, 

particularly of spiked samples or from loss of radioactive standards containing heavy water, 

radioiodines, or noble gases. 

4.2.1.1.7 Radioactive Waste Management Sites 

Areas 3 and 5 contain sites for the disposal of low-level radioactive waste, and Area 5 contains sites 

for storage of transuranic and mixed transuranic wastes, as well as the Greater Confinement Disposal 

Test Unit and 12 accompanying boreholes (only a few contain any waste). Disposal occurs in pits 

and trenches; concrete pads provide temporary storage of certain wastes. Area 5 is for packaged 

wastes disposal only. The Waste Examination Facility (WEF) houses a HEPA-filtered glovebox used 

to examine and repack TRU waste drums. No contamination has been released from glove box 

operations to the environment. The drums, sent to the NTS from LLNL in years past, are stored inside 

the TRU Pad Cover Building. Repacked drums will be sent to the WIPP. The facility is a diffuse 

source of radiological effluents. The only radioactive effluent detected by the various types of 

samplers surrounding the site is HTO in atmospheric moisture. The Area 3 low-level waste site is in a 

location where surface soil has been contaminated by deposited plutonium, and resuspension of this 

soil by wind and vehicular activity has resulted in above-background levels of plutonium being 

detected in air samples collected nearby.
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4.2.1.2 Atmospheric Radionuclide Concentrations

In 1964, REECo established an environmental surveillance program at NTS designed to measure 

radiological conditions throughout the site without regard to nuclear tests. That is, the data collected 

by the program was not to relate to specific tests but to general conditions of radiation. The 

short-term objective of the program was to minimize casual personnel exposure to radiation by 

locating and identifying localized radiological environmental conditions by type and quantity of 

contamination. The long-range objective of the program was to establish baseline environmental data 

that could provide a reference for comparison with subsequent test activities and radiation 

measurements.

The initial surveillance program included, over time, 12 permanent air-sampling stations in the most 

populated areas at NTS. The air samplers were low-volume Filter Queen samplers with 8- by 10-in. 

(Gelman Type E) glass-fiber filter papers. Operating times were determined by integrated electric 

timers with flow rates determined by calibrated rotometers. Typical flow rates varied from 3 to 6 cubic 

feet per minute; samples were collected weekly.

After the first reporting period (June 1964), positive-displacement Gast pumps, which were equipped 

with in-line total volume gas meters, replaced the Filter Queen samplers. In May and June of 1965, 

the 8- by 10-in. Gelman filters were replaced with a new sampler that used 4-in.-diameter 

Whatman #41 filter paper. The sampling rate of these samplers was about 3 cubic feet per minute. 

Therefore, the total volume of sampled air in a 7-day period was about 1 × 103

 m3. During this period, 

the number of sampling stations increased to 13, and caustic scrubbers were added for the detection 

of radioiodines.

Early particulate samples were typically analyzed only for gross alpha and gross beta. However, if 

gross beta concentrations exceeded 1 × 10-5 μCi m-3, researchers conducted an analysis for 227Ac (the 

most hazardous beta emitter present). Because no historical evidence exists that 227Ac has been

detected in air or soil samples, the assumption that unidentified beta emitters were 227Ac would be 

unreasonable and inappropriate. Therefore, for purposes of dose reconstruction, 90Sr is assumed to 

be the unidentified beta emitting radionuclide with the highest internal dose impact (see Section 

4.2.1.2.4 for details).

Environmental samples were analyzed for gross alpha and beta radioactivity by gas proportional 

counting. Air samples were analyzed with a Nuclear Chicago ULTRASCALER system having a 

9- by 10-in. detection chamber to accommodate the 8- by 10-in. filters. All other samples were 

analyzed with a Beckman WIDEBETA system equipped with an automatic sample changer. This 

counting equipment and protocol allowed lower detection limits of about 2.9 × 10-9 and 

1.6 × 10-8 μCi m-3 for gross alpha and gross beta, respectively (Lewis, Glora, and Aoki 1965). By 

1978, the lower detection limit for gross beta counting had been reduced to 1 × 10-10 μCi m-3. 

Gross gamma screening was typically performed on samples prepared for gross beta counting. The 

screening was performed using a 5- by 5-in. NaI(Tl) detector coupled to a single-channel analyzer. 

Any samples showing elevated gamma readings were sent to a multichannel analyzer for radionuclide 

identification. In the early 1970s, gamma spectroscopy was performed using germanium–lithium 

[Ge(Li)] detectors and 2,000-channel analyzers with a lower detection limit of 5 × 10-9 μCi m-3. After 

cessation of atmospheric testing at NTS in 1963, fission products were frequently measured in the 

atmosphere but were typically correlated with foreign atmospheric weapons testing (see Section 

4.2.1.2.4). 
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In 1971, weekly air samples from a given station were batched on a monthly basis and subjected to 

radiochemical analysis for 239Pu. The procedure included acid dissolution with ion exchange recovery 

and electroplating onto stainless-steel discs, followed by alpha spectroscopy using solid-state surface 

barrier detectors. This analysis provided a nominal minimum detection limit of 3 × 10-7 μCi m-3 (Lantz

1978a). Routine analysis for 238Pu started in 1989.

In 1977, a separate sampler was designed for the collection of airborne tritium (HT) and tritiated water 

vapor (HTO). The sampler was portable and capable of unattended operation for up to 2 weeks in 

desert areas. A small electronic pump drew air into the sampler at about 0.5 liter per minute, and the 

HTO was removed from the air stream by a silica gel drying column. The dry air then passed through 

a catalytic converter containing platinum to generate HTO from HT. Another drying column collected 

the vapor, in which a small volume of distilled water served as a trap for the HTO. Appropriate 

aliquots of condensed moisture were obtained by heating the silica gel. Counting via liquid 

scintillation techniques allowed for the determination of the HT and HTO activities. The typical 

minimum detection limit for this analysis was 3 × 10-7 μCi m-3. 

The total number of air sampling stations increased over the years to a peak of 52 in 1989. This 

number remained fairly constant until a gradual reduction began in 1995. This reduction occurred 

primarily because of a gradual strategy shift from environmental monitoring to demonstration of 

compliance with National Emission Standards for Hazardous Air Pollutants as approved by the U.S. 

Environmental Protection Agency (EPA). Figure 4.2.1.2-1 shows the locations of the 48 sampling 

stations in 1997.

Since the early 1970s, the environmental surveillance program has routinely monitored atmospheric 

concentrations of tritium and 238Pu, 239Pu, and 240Pu. These were the radionuclides considered most 

important to dose for workers and members of the public. In addition, since the mid-1960s, 

measurements have been reported for gross alpha and gross beta concentrations. The following 

sections discuss the history of these measurements in NTS annual environmental reports (see 

reference section for a listing) and their importance to dose reconstruction for unmonitored 

employees.

4.2.1.2.1 Tritium

Atmospheric measurements at NTS began in 1977 with samplers in Areas 5, 10, and 23. The 

sampler near Building 650 (Mercury) was designated as the control station. During that year, the 

samples in Area 10 (near the Sedan Crater) demonstrated the highest tritium (HTO) concentrations 

with a high of 3.0 × 10-4 pCi m-3. The number of sampling locations increased over the following years 

to include all the areas listed in Table 4.2.1.2.1-1. These locations provided representative samples 

from the most populated areas on the site.

The tritium concentrations in Table 4.2.1.2.1-1 are typically the average of the maximum 

concentrations for a given area in a given year, except when these values were determined to be 

adversely affected by nearby point-source releases (e.g., Building 790 in 1983; Scoggins 1983, p. 29). 

In cases where maximum values were not provided, the average concentrations were reported. 

In addition to the annual area tritium concentrations, Table 4.2.1.2.1-1 provides NTS site average and 

maximum concentrations, which represent the arithmetic averages of the concentrations of all the 

areas and the maximum of all the areas, respectively. For purposes of dose reconstruction, these 

average and maximum site concentrations have been converted to annual organ dose using the dose 

conversion factor for inhalation of 1.80 × 10-11 Sv Bq-1 (ICRP 1998) and by assuming submersion in 

the respective concentrations for 2,000 hr yr-1 and an inhalation rate of 2,400 m3 yr-1. 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 12 of 51

Figure 4.2.1.2-1. Air sampling stations in 1997. 

In addition to the inhalation dose, a factor of 1.5 was included in the organ dose calculation to account 

for absorption through the skin.

The uncertainty associated with the values in Table 4.2.1.2.1-1 were not provided in the annual 

environmental report until 2001. Based on the values provided for that year, the standard deviation 

would be typically less than 10% of the measured values. Because sampling methods were changed 
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Table 4.2.1.2.1-1. Historical HTO atmospheric concentrations (pCi ml-1) by year and Area with estimated maximum and average 

annual organ dose. 

Year

Area at NTS

Site 

average

(pCi ml1

)

Site 

maximum

(pCi ml-1)

Site 

maximum 

annual 

dose

(mrem)

Site 

average 

annual 

dose

1 3 5 6 10 12 15 16 18 20 23 25 (mrem)

1977 3.0E-04 1.2E-06 1.5E-04 3.0E-04 6.9E-02 3.5E-02

1978 2.9E-03 1.3E-04 1.9E-05 1.9E-05 3.0E-05 6.2E-04 2.9E-03 6.7E-01 1.4E-01

1979 6.2E-04a 2.9E-03 a 6.7E-01 a 1.4E-01 a

1980 2.6E-04 7.1E-05 1.7E-04 2.6E-04 6.0E-02 3.8E-02

1981 3.6E-03 9.8E-04 2.3E-03 3.6E-03 8.3E-01 5.3E-01

1982 7.5E-04 2.3E-03 4.4E-03 7.6E-04 5.0E-03 2.6E-03 5.0E-03 1.2E+00 6.1E-01

1983 8.2E-05 1.1E-04 8.9E-05 3.9E-03 6.5E-04 9.6E-04 3.9E-03 8.9E-01 2.2E-01

1984 1.7E-04 5.3E-04 3.3E-05 3.5E-05 5.0E-03 1.2E-03 5.0E-03 1.2E+00 2.7E-01

1985 1.4E-04 8.6E-05 4.7E-03 3.5E-05 4.7E-04 1.1E-03 4.7E-03 1.1E+00 2.5E-01

1986 1.2E-04 5.6E-05 6.6E-05 6.0E-05 8.4E-04 2.3E-04 8.4E-04 1.9E-01 5.3E-02

1987 4.5E-05 1.7E-04 6.3E-05 4.6E-04 2.6E-04 2.0E-04 4.6E-04 1.0E-01 4.6E-02

1988 4.4E-05 3.9E-05 2.0E-05 5.2E-04 3.2E-05 1.3E-04 5.2E-04 1.2E-01 3.0E-02

1989 1.7E-04 3.1E-05 1.3E-05 1.9E-05 8.2E-05 4.6E-06 5.3E-05 1.7E-04 3.9E-02 1.2E-02

1990 7.4E-06 2.2E-05 1.3E-05 4.0E-06 2.3E-05 7.5E-05

2.1E05 2.4E-05 7.5E-05 1.7E-02 5.5E-03

1991 9.1E-06 2.8E-05 6.3E-06 8.4E-06 1.7E-05 4.3E-06

2.1E06 1.1E-05 2.8E-05 6.5E-03 2.5E-03

1992 1.0E-05 3.5E-05 3.3E-06 6.2E-05 6.2E-05 6.1E-06

8.8E05 3.8E-05 8.8E-05 2.0E-02 8.8E-03

1993 8.7E-06 2.8E-05 7.9E-06 3.7E-06 2.0E-05 4.0E-06

1.9E06 1.1E-05 2.8E-05 6.5E-03 2.4E-03

1994

1995 3.7E-06 1.4E-06 2.1E-05 9.5E-07 7.0E-06 4.3E-06 1.0E-05 2.6E-06

1.4E06 5.8E-06 2.1E-05 4.9E-03 1.3E-03

1996 3.1E-06 4.1E-06 1.6E-05 5.6E-05 2.6E-05 2.0E-05 1.4E-05 1.1E-06

1.1E06 1.6E-05 5.6E-05 1.3E-02 3.6E-03

1997 2.2E-06 3.8E-06 7.5E-06 5.6E-05 2.9E-05 2.3E-05 2.0E-05 1.3E-06

1.3E06 1.6E-05 5.6E-05 1.3E-02 3.7E-03

1998 3.4E-06 1.7E-05 1.8E-04 2.9E-05 5.5E-05 1.4E-05 4.6E-04 1.1E-04 4.6E-04 1.1E-01 2.5E-02

1999 2.0E-05 1.8E-05 4.2E-06 4.1E-05 2.8E-05 2.7E-05 7.5E-04 1.3E-04 7.5E-04 1.7E-01 2.9E-02

2000 4.1E-06 1.5E-05 5.1E-05 5.4E-05 1.9E-05 9.7E-04 1.9E-04 9.7E-04 2.2E-01 4.3E-02

2001 4.2E-06 3.6E-06 6.2E-06 4.2E-05 2.5E-05 3.8E-06 2.3E-06 1.3E-06 1.1E-03 7.1E-07

6.1E06 1.1E-04 1.1E-03 2.5E-01 2.5E-02

a. Value not provided in annual reports - assumed same as for 1978.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 14 of 51

over time and the unavailability of uncertainty estimates for previous years, the favorable to claimant

assumption is made that the 50th-percentile expected intakes are those in Table 4.2.1.2.1-1 and that 

the 95th-percentile values would be enveloped by a factor of ±2 (i.e., ±100%).

For purposes of dose reconstruction, the geometric deviation (GSD) of the values in Table 4.2.1.2.2-1 

would be estimated by using the following expression for a log-normal distribution:

)()( .655851

1

 50th percentile

 95th percentile GSD = (4-1)

If the assumption is made that the 95th-percentile values would be included in a 2 sigma of ±100%, 

then the geometric standard deviation can be shown using equation 4-1 to be 1.52.

If the favorable to claimant assumption is used to estimate the 95th-percentile dose (i.e., add 100%) 

from the maximum annual doses in Table 4.2.1.2.1-1, the resultant annual organ doses would still be 

less than 2 mrem yr-1. For purposes of dose reconstruction, these small doses would be 

inconsequential to estimates of causation and should therefore not be included in dose 

reconstructions of unmonitored employees. Also, because these maximum annual dose estimates 

were based on empirical measurements at locations of known sources of tritium emission, it is 

considered unlikely that average ambient atmospheric concentrations of tritium prior to 1977 (when 

data are not readily available), dose reconstructors should not assign missed environmental tritium 

dose for any unmonitored covered employee.

4.2.1.2.2 Plutonium

Routine isotopic atmospheric measurements of plutonium at NTS began in 1971 with samplers in 15 

locations across the site. Six additional sampling stations were added in 1978. These locations were 

chosen to provide representative samples from the populated areas on the site as well as to monitor 

resuspension of low-fired plutonium spread by safety experiments before 1960 in Areas 2, 3, 4, 7, 9, 

and 10.

The 239Pu and 238Pu concentrations in Tables 4.2.1.2.2-1 and 4.2.1.2.2-3, respectively, are typically 

the average of the maximum concentrations for a given area in a given year. In cases where 

maximum values were not provided, the average of the average concentrations was reported. In 

addition to the annual area concentrations, Tables 4.2.1.2.2-1 and 4.2.1.2.2-3 provide NTS site 

average and maximum concentrations, which represent the arithmetic averages of the concentration 

of all the areas and the maximum of all the areas, respectively. Potential intakes associated with 

these concentrations can be calculated under the assumption that an unmonitored worker was 

occupationally exposed for 2,000 hr yr-1 with a breathing rate of 2,400 m3 yr-1. Tables 4.2.1.2.2-2 and 

4.2.1.2.2-4 present these estimated intakes.

It should also be noted that some covered employees remained onsite continuously for weeks at a 

time. However, because most of the non-working hours were spent indoors where ambient air 

particulate loadings would be greatly reduced from the outdoor loadings and because of the 

conservative assumptions used to estimate the values in Tables 4.2.1.2.2-2 and 4.2.1.2.2-4, 

adjustment of the tabular data is not required to assure intakes are not underestimated for these 

individuals.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 15 of 51

Table 4.2.1.2.2-1. 239Pu atmospheric concentrations (pCi m-3) for sampled Areas. 

Area

Year 1 2 3 4 5 6 7 9 10 11 12

1971 3.7E-04 6.6E-04 1.7E-04 1.2E-04 1.7E-04 7.2E-04 2.1E-04 9.1E-05 1.6E-04

1972 1.5E-04 2.2E-04 3.7E-04 1.4E-04 1.3E-03 4.3E-03 2.9E-04 2.4E-04 7.9E-04

1973 8.7E-05 2.2E-04 2.1E-04 8.7E-05 1.2E-04 8.6E-04 4.4E-05 2.4E-04 4.3E-05

1974 8.0E-05 6.8E-05 1.3E-04 5.0E-05 7.9E-05 2.1E-04 5.7E-05 8.0E-05 5.8E-05

1975 2.9E-05 3.8E-05 1.5E-04 4.3E-05 6.3E-05 1.7E-04 5.7E-05 4.9E-05 3.3E-05

1976 2.9E-04 1.7E-04 1.0E-03 1.5E-04 4.4E-04 3.2E-03 2.1E-04 2.4E-04 2.8E-04

1977 2.7E-05 6.5E-05 2.6E-04 3.4E-05 5.5E-05 2.5E-04 4.6E-05 1.2E-04 2.1E-05

1978 4.7E-05 1.3E-04 2.8E-04 6.1E-05 9.9E-05 5.4E-04 2.9E-04 1.1E-04 6.2E-05

1979 7.4E-05 4.7E-04 1.7E-06 2.1E-05 4.0E-05 4.9E-05 5.2E-04 2.9E-04 3.7E-05 2.4E-05

1980 2.3E-05 2.6E-04 1.1E-06 2.9E-05 4.8E-05 4.5E-05 4.5E-04 4.4E-05 1.9E-05

1981 2.6E-05 6.4E-05 1.4E-04 2.5E-05 2.9E-05 3.3E-05 3.2E-04 4.6E-05 2.0E-05

1982 6.1E-05 4.0E-05 6.2E-05 2.0E-05 3.5E-05 7.4E-05 2.2E-04 5.3E-05 2.7E-05

1983 3.1E-05 6.5E-05 9.2E-05 2.3E-05 3.2E-05 3.3E-05 2.1E-04 1.8E-04 2.3E-05

1984 1.9E-05 1.0E-04 2.3E-04 3.8E-05 2.0E-05 3.8E-05 1.0E-03 5.6E-05 1.7E-05

1985 1.5E-05 5.8E-05 2.1E-04 3.1E-05 2.7E-05 2.7E-05 1.3E-03 5.2E-05 2.3E-05

1986 1.5E-04 6.2E-05 5.5E-04 2.8E-05 1.9E-04 4.8E-05 2.8E-04 2.5E-05 3.2E-05

1987 5.5E-05 2.3E-05 2.8E-03 1.7E-05 2.5E-05 1.6E-05 1.1E-04 3.2E-05 1.6E-05

1988 9.2E-05 2.8E-05 2.1E-04 2.8E-05 2.2E-05 1.5E-05 5.1E-05 1.2E-05 1.6E-05

1989 8.6E-04 5.8E-05 5.4E-04 5.0E-05 8.0E-05 2.7E-04 3.5E-04 5.4E-04 6.8E-04

1990 9.9E-05 3.2E-05 3.5E-04 3.1E-05 3.7E-05 4.1E-05 5.0E-04 4.3E-05 1.1E-04 1.0E-05

1991 3.1E-03 2.1E-05 3.8E-04 3.5E-05 7.2E-05 3.2E-05 3.5E-04 5.4E-07 2.2E-04 1.4E-05

1992 1.5E-04 1.1E-04 2.3E-03 8.0E-05 7.3E-05 3.9E-04 8.8E-04 2.4E-04 8.8E-05 3.8E-05

1993 2.0E-04 4.9E-05 4.0E-04 8.0E-05 9.0E-05 7.8E-05 2.7E-03 1.7E-04 4.2E-04 8.4E-04

1994

1995 4.1E-05 1.1E-05 2.4E-04 1.5E-05 5.0E-05 1.2E-05 3.9E-04 3.8E-05 4.0E-05 1.1E-05

1996 7.5E-04 3.0E-04 1.6E-05 4.6E-05 4.5E-04 6.1E-04 5.8E-05 1.9E-05 3.7E-06

1997 1.1E-05 2.6E-05 1.1E-05 1.5E-06

1998 4.7E-04 6.7E-05 9.5E-05 1.1E-05 2.7E-05 7.4E-04 9.5E-06

1999 5.9E-05

2000 8.5E-05 2.7E-04 5.9E-05 2.1E-05 4.1E-04 3.4E-05 2.8E-03

2001 4.5E-04 1.2E-04 1.2E-04 1.9E-04 6.8E-05 9.6E-05 1.7E-05 5.0E-04 8.5E-06
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 16 of 51

Table 4.2.1.2.2-1 (cont.). 239Pu atmospheric concentrations (pCi m-3) for sampled Areas. 

Year

Area

15 16 18 19 20 23 25 27 28 Site average Site maximum

1971 1.9E-04 8.0E-05 2.3E-04 9.4E-05 8.4E-05 3.9E-05 2.26E-04 7.21E-04

1972 1.7E-04 1.5E-04 2.7E-03 3.5E-04 1.7E-04 5.1E-05 7.63E-04 4.29E-03

1973 4.4E-05 3.8E-05 7.2E-05 5.6E-05 3.7E-05 3.3E-05 1.46E-04 8.59E-04

1974 6.0E-05 6.6E-05 8.5E-05 8.6E-05 7.2E-05 1.2E-04 8.64E-05 2.11E-04

1975 4.3E-05 3.4E-05 4.4E-05 5.7E-05 6.8E-05 7.0E-05 6.27E-05 1.68E-04

1976 1.6E-04 9.9E-05 5.7E-04 2.0E-04 1.4E-04 6.3E-05 4.82E-04 3.18E-03

1977 2.6E-05 2.3E-05 2.7E-05 3.3E-05 3.1E-05 2.3E-05 6.93E-05 2.55E-04

1978 5.1E-05 6.2E-05 5.8E-05 7.0E-05 9.4E-05 3.9E-05 1.33E-04 5.42E-04

1979 5.2E-05 2.1E-05 2.0E-05 2.2E-05 1.8E-05 1.6E-05 2.3E-05 9.98E-05 5.24E-04

1980 6.4E-05 9.5E-05 3.5E-05 3.1E-05 2.1E-05 1.6E-05 7.88E-05 4.52E-04

1981 9.3E-05 2.6E-05 2.3E-05 2.9E-05 2.4E-05 2.5E-05 1.6E-05 5.88E-05 3.22E-04

1982 4.2E-05 1.7E-05 2.2E-05 2.1E-05 1.7E-05 5.04E-05 2.15E-04

1983 2.2E-05 1.5E-05 3.0E-05 2.8E-05 3.0E-05 5.85E-05 2.14E-04

1984 3.2E-05 2.8E-05 2.5E-05 2.9E-05 3.2E-05 4.8E-05 1.15E-04 1.02E-03

1985 4.0E-05 2.1E-05 3.5E-05 2.2E-05 2.6E-05 3.1E-05 1.30E-04 1.33E-03

1986 3.9E-05 2.8E-05 2.2E-05 6.6E-05 2.1E-05 2.3E-05 1.03E-04 5.45E-04

1987 1.8E-05 1.4E-05 1.7E-05 1.6E-05 2.3E-05 2.13E-04 2.81E-03

1988 2.4E-05 1.4E-05 3.5E-05 1.5E-05 1.5E-05 4.15E-05 2.14E-04

1989 2.3E-05 1.1E-05 9.0E-06 1.2E-04 9.5E-06 4.8E-06 2.40E-04 8.55E-04

1990 8.2E-05 5.8E-06 3.9E-06 8.5E-05 1.1E-05 3.5E-05 9.17E-05 4.98E-04

1991 1.2E-04 2.9E-05 8.8E-06 1.3E-05 2.6E-05 7.7E-06 2.77E-04 3.10E-03

1992 1.1E-03 1.9E-05 8.9E-06 1.3E-05 1.8E-05 1.6E-04 3.52E-04 2.26E-03

1993 3.3E-04 1.3E-05 7.7E-06 6.3E-06 3.4E-06 1.1E-05 3.37E-04 2.70E-03

1994

1995 1.6E-04 1.4E-05 1.0E-06 7.4E-06 4.5E-06 6.92E-05 3.90E-04

1996 2.4E-04 1.8E-06 3.5E-06 4.5E-06 6.3E-06 4.2E-06 2.6E-06 1.57E-04 7.50E-04

1997 2.9E-05 1.9E-07 2.0E-05 2.0E-05 6.0E-07 1.0E-06 1.6E-06 1.11E-05 2.90E-05

1998 6.3E-05 3.1E-05 3.3E-05 1.2E-05 6.3E-06 1.30E-04 7.35E-04

1999 1.1E-05 1.1E-05 2.0E-05 6.2E-06 2.14E-05 5.90E-05

2000 2.6E-04 1.0E-05 6.6E-06 1.1E-05 3.63E-04 2.83E-03

2001 1.4E-05 9.6E-06 1.7E-05 7.6E-06 2.0E-06 4.4E-06 1.02E-04 5.04E-04
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 17 of 51

Table 4.2.1.2.2-2. Annual 239Pu intakes (Bq) for sampled Areas. 

Year

Area

12345679 10 11 12

1971 0.0329 0.0588 0.0147 0.0105 0.0153 0.0641 0.0185 0.0081 0.0145

1972 0.0132 0.0199 0.03270.0124 0.1191 0.3813 0.0258 0.0216 0.0699

1973 0.0077 0.0198 0.0183 0.0078 0.0107 0.0764 0.0039 0.0216 0.0038

1974 0.0071 0.0060 0.0112 0.0044 0.0070 0.0188 0.0051 0.0071 0.0052

1975 0.0026 0.0034 0.0130 0.0038 0.0056 0.0149 0.0051 0.0044 0.0029

1976 0.0255 0.0153 0.0924 0.0133 0.0393 0.2827 0.0189 0.0210 0.0248

1977 0.0024 0.0058 0.0227 0.0030 0.0049 0.0225 0.0041 0.0108 0.0019

1978 0.0042 0.0118 0.0251 0.0054 0.0088 0.0482 0.0258 0.0095 0.0055

1979 0.0066 0.0413 0.0001 0.0018 0.0035 0.0044 0.0466 0.0258 0.0033 0.0021

1980 0.0020 0.0230 0.0001 0.0026 0.0043 0.0040 0.0402 0.0039 0.0017

1981 0.0023 0.0056 0.0124 0.0022 0.0026 0.0029 0.0286 0.0041 0.0018

1982 0.0054 0.0036 0.0055 0.0018 0.0031 0.0066 0.0191 0.0047 0.0024

1983 0.0028 0.0057 0.0082 0.0020 0.0029 0.0029 0.0190 0.0162 0.0020

1984 0.0017 0.0088 0.0203 0.0034 0.0018 0.0034 0.0907 0.0050 0.0015

1985 0.0013 0.0051 0.0189 0.0027 0.0024 0.0024 0.1182 0.0046 0.0020

1986 0.0132 0.0055 0.0484 0.0025 0.0165 0.0043 0.0249 0.0022 0.0028

1987 0.0048 0.0020 0.2498 0.0015 0.0022 0.0014 0.0098 0.0028 0.0014

1988 0.0082 0.0025 0.0190 0.0025 0.0020 0.0013 0.0045 0.0011 0.0014

1989 0.0760 0.0052 0.0478 0.0045 0.0071 0.0240 0.0311 0.0480 0.0604

1990 0.0088 0.0029 0.0308 0.0027 0.0033 0.0036 0.0443 0.0038 0.0095 0.0009

1991 0.2756 0.0018 0.0340 0.0031 0.0064 0.0029 0.0311 0.0000 0.0199 0.0013

1992 0.0133 0.0098 0.2009 0.0071 0.0065 0.0347 0.0782 0.0213 0.0078 0.0034

1993 0.0180 0.0044 0.0352 0.0071 0.0080 0.00690.2400 0.0151 0.0373 0.0747

1994

1995 0.0036 0.0010 0.0216 0.0014 0.0045 0.0011 0.0347 0.0033 0.0036 0.0010

1996 0.0667 0.0268 0.0015 0.0041 0.0400 0.0542 0.0051 0.0017 0.0003

1997 0.0010 0.0023 0.0010 0.0001

1998 0.0417 0.00600.0085 0.0010 0.0024 0.0653 0.0008

1999 0.0052

2000 0.0075 0.0239 0.0052 0.0018 0.0366 0.0030 0.2511

2001 0.0404 0.0011 0.0103 0.0166 0.0061 0.0086 0.0015 0.0448 0.0008
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 18 of 51

Table 4.2.1.2.2-2 (cont.). Annual 239Pu intakes (Bq) for sampled Areas.

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 0.0172 0.0071 0.0201 0.0083 0.0075 0.0035 0.0201 0.0201 0.0641

1972 0.0150 0.0129 0.2427 0.0309 0.0148 0.0045 0.0678 0.0678 0.3813

1973 0.0039 0.0034 0.0064 0.0050 0.0033 0.0029 0.0130 0.0130 0.0764

1974 0.0053 0.0059 0.0076 0.0076 0.0064 0.0104 0.0077 0.0077 0.0188

1975 0.0038 0.0030 0.0039 0.0051 0.0060 0.0062 0.0056 0.0056 0.0149

1976 0.0139 0.0088 0.0505 0.0179 0.0122 0.0056 0.0428 0.0428 0.2827

1977 0.0023 0.0020 0.0024 0.0029 0.0276 0.0020 0.0078 0.0078 0.0276

1978 0.0045 0.0055 0.0052 0.0062 0.0084 0.0035 0.0118 0.0118 0.0482

1979 0.0019 0.0017 0.0020 0.0016 0.0014 0.0020 0.0089 0.0089 0.0466

1980 0.0085 0.0031 0.0027 0.0019 0.0014 0.0070 0.0070 0.0402

1981 0.0023 0.0020 0.0026 0.0022 0.0022 0.0014 0.0052 0.0052 0.0286

1982 0.0015 0.0019 0.0019 0.0015 0.0045 0.0045 0.0191

1983 0.0013 0.0026 0.0025 0.0027 0.0052 0.0052 0.0190

1984 0.0025 0.0022 0.0026 0.0028 0.0043 0.0102 0.0102 0.0907

1985 0.0019 0.0031 0.0020 0.0023 0.0028 0.0116 0.0116 0.1182

1986 0.0025 0.0020 0.0058 0.0018 0.0020 0.0092 0.0092 0.0484

1987 0.0012 0.0015 0.0014 0.0020 0.0203 0.0203 0.2498

1988 0.0000 0.0031 0.0013 0.0013 0.0036 0.0036 0.0190

1989 0.0010 0.0008 0.0104 0.0008 0.0004 0.0213 0.0213 0.0760

1990 0.0005 0.0003 0.0075 0.0010 0.0031 0.0082 0.0082 0.0443

1991 0.0026 0.0008 0.0012 0.0023 0.0007 0.0247 0.0247 0.2756

1992 0.0017 0.0008 0.0011 0.0016 0.0142 0.0313 0.0313 0.2009

1993 0.0012 0.0007 0.0006 0.0003 0.0010 0.0300 0.0300 0.2400

1994

1995 0.0012 0.0001 0.0007 0.0004 0.0061 0.0061 0.0347

1996 0.0002 0.0003 0.0004 0.0006 0.0004 0.0002 0.0140 0.0140 0.0667

1997 0.0000 0.0018 0.0018 0.0001 0.0001 0.0001 0.0010 0.0010 0.0026

1998 0.0027 0.0029 0.0010 0.0006 0.0115 0.0115 0.0653

1999 0.0010 0.0018 0.0006 0.0019 0.0019 0.0052

2000 0.0009 0.0006 0.0010 0.0323 0.0323 0.2511

2001 0.0008 0.0015 0.0007 0.0002 0.0004 0.0090 0.0090 0.0448
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 19 of 51

Table 4.2.1.2.2-3. 238Pu atmospheric concentrations (pCi m-3) for sampled Areas. 

Year

Area

1 2 3 4 5 6 7 9 10 11 12

1989 2.15E-05 1.60E-05 1.69E-05 1.43E-05 1.21E-05 2.60E-05 1.60E-05 0.00E+00 1.40E-05 2.00E-05

1990 2.23E-05 2.15E-05 1.43E-05 1.26E-05 1.40E-05 1.54E-05 1.99E-05 1.75E-05 1.21E-05 1.72E-05

1991 1.33E-05 1.30E-05 1.39E-05 1.12E-05 7.77E-06 1.46E-05 2.43E-05 1.53E-05 1.44E-05 4.50E-06

1992 1.23E-05 3.95E-06 7.80E-05 8.09E-06 6.02E-06 3.50E-06 1.30E-05 5.00E-06 5.60E-06 1.40E-05

1993 5.25E-06 2.64E-05 8.51E-06 4.18E-06 3.33E-06 1.40E-05 4.80E-05 7.60E-06 7.80E-06 9.70E-06

1994 1.23E-04 4.48E-06 1.08E-05 2.18E-06 1.92E-06 1.90E-06 2.60E-05 2.40E-06 3.50E-06 1.20E-06

1995 5.45E-06 1.90E-06 4.08E-06 1.51E-06 3.37E-06 8.10E-06 4.60E-06 1.30E-06 5.30E-07 9.50E-07

1996 7.60E-06 1.49E-06 8.07E-06 1.86E-06 1.52E-06 5.00E-06 9.20E-06 9.50E-07 2.50E-07 1.00E-06

1997 7.90E-07 4.00E-07 3.75E-07 6.90E-06 1.03E-07 4.47E-07 7.60E-08 4.20E-06 1.15E-06 3.90E-08 6.00E-07

1998 1.70E-06 2.85E-07 1.25E-06 3.52E-07 1.40E-07 3.30E-06 3.20E-06 2.90E-07

1999 2.70E-07 2.60E-07 7.95E-07 8.30E-06 2.30E-07 1.40E-05 3.50E-06

2000 9.89E-06 9.65E-06 8.11E-06 2.91E-05 8.47E-06 9.05E-06 5.41E-06 4.67E-05 1.27E-05

2001 1.12E-05 5.54E-06 1.11E-05 4.62E-05 5.96E-06 9.15E-06 4.78E-06 8.36E-06 1.40E-05

Table 4.2.1.2.2-3 (cont.). 238Pu atmospheric concentrations (pCi m-3) for areas sampled.

Year

Area Site 

average

Site 

13 15 16 18 19 20 23 25 27 maximum

1989 2.07E-05 3.90E-05 1.75E-05 1.80E-05 1.55E-05 2.90E-05 2.20E-05 1.9E-05 3.9E-05

1990 1.36E-05 1.13E-05 1.80E-05 1.02E-05 1.63E-05 1.39E-05 4.62E-06 1.5E-05 2.2E-05

1991 1.09E-05 1.17E-05 1.77E-05 1.41E-05 1.27E-05 1.45E-05 8.04E-06 1.3E-05 2.4E-05

1992 1.50E-05 1.90E-06 6.80E-06 1.12E-06 2.95E-06 9.60E-06 1.2E-05 7.8E-05

1993 7.20E-06 7.70E-07 1.90E-06 1.10E-06 1.24E-06 3.05E-06 1.70E-06 8.9E-06 4.8E-05

1994 8.80E-06 9.60E-07 1.6E-05 1.2E-04

1995 5.40E-06 2.75E-07 3.50E-06 3.53E-07 5.10E-07 2.8E-06 8.1E-06

1996 3.10E-06 4.70E-06 2.10E-07 1.90E-07 3.90E-06 1.45E-07 7.80E-07 1.10E-06 2.8E-06 9.2E-06

1997 7.70E-07 3.00E-07 1.50E-08 1.50E-07 1.90E-08 6.80E-08 9.6E-07 6.9E-06

1998 4.10E-07 2.50E-07 1.90E-07 1.0E-06 3.3E-06

1999 1.93E-06 3.7E-06 1.4E-05

2000 5.64E-06 6.62E-06 5.66E-06 1.09E-05 1.3E-05 4.7E-05

2001 7.59E-06 2.12E-06 1.09E-05 7.58E-06 2.58E-06 4.77E-06 1.0E-05 4.6E-05
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 20 of 51

Table 4.2.1.2.2-4. 238Pu intakes (Bq) for sampled Areas. 

Year

Area

12345679 10 11 12

1989 1.9E-03 1.4E-03 1.5E-03 1.3E-03 1.1E-03 2.3E-03 1.4E-03 1.2E-03 1.8E-03

1990 2.0E-03 1.9E-03 1.3E-03 1.1E-03 1.2E-03 1.4E-03 1.8E-03 1.6E-03 1.1E-03 1.5E-03

1991 1.2E-03 1.2E-03 1.2E-03 1.0E-03 6.9E-04 1.3E-03 2.2E-03 1.4E-03 1.3E-03 4.0E-04

1992 1.1E-03 3.5E-04 6.9E-03 7.2E-04 5.4E-04 3.1E-04 1.2E-03 4.4E-04 5.0E-04 1.2E-03

1993 4.7E-04 2.3E-03 7.6E-04 3.7E-04 3.0E-04 1.2E-03 4.3E-03 6.8E-04 6.9E-04 8.6E-04

1994 1.1E-02 4.0E-04 9.6E-04 1.9E-04 1.7E-04 1.7E-04 2.3E-03 2.1E-04 3.1E-04 1.1E-04

1995 4.8E-04 1.7E-04 3.6E-04 1.3E-04 3.0E-04 7.2E-04 4.1E-04 1.2E-04 4.7E-05 8.4E-05

1996 6.8E-04 1.3E-04 7.2E-041.7E-04 1.4E-04 4.4E-04 8.2E-04 8.4E-05 2.2E-05 8.9E-05

1997 7.0E-05 3.6E-05 3.3E-05 6.1E-04 9.2E-06 4.0E-05 6.8E-06 3.7E-04 1.0E-04 3.5E-06 5.3E-05

1998 1.5E-04 2.5E-05 1.1E-04 3.1E-05 1.2E-05 2.9E-04 2.8E-04 2.6E-05

1999 2.4E-05 2.3E-05 7.1E-05 7.4E-04 2.0E-05 1.2E-03 3.1E-04

2000 8.8E-04 8.6E-04 7.2E-04 2.6E-03 7.5E-04 8.0E-04 4.8E-04 4.2E-03 1.1E-03

2001 1.0E-03 4.9E-04 9.9E-04 4.1E-03 5.3E-04 8.1E-04 4.2E-04 7.4E-04 1.2E-03

Table 4.2.1.2.2-4 (cont.). 238Pu intakes (Bq) for sampled Areas.

Year

Area Site 

average

Site 

13 15 16 18 19 20 23 25 27 maximum

1989 1.8E-03 3.5E-03 1.6E-03 1.6E-03 1.4E-03 2.6E-03 2.0E-03 1.7E-03 3.5E-03

1990 1.2E-03 1.0E-03 1.6E-03 9.1E-04 1.4E-03 1.2E-03 4.1E-04 1.3E-03 2.0E-03

1991 9.7E-04 1.0E-031.6E-03 1.3E-03 1.1E-03 1.3E-03 7.1E-04 1.2E-03 2.2E-03

1992 1.3E-03 1.7E-04 6.0E-04 1.0E-04 2.6E-04 8.5E-04 1.0E-03 6.9E-03

1993 6.4E-04 6.8E-05 1.7E-04 9.8E-05 1.1E-04 2.7E-04 1.5E-04 7.9E-04 4.3E-03

1994 7.8E-04 8.5E-05 1.4E-03 1.1E-02

1995 4.8E-04 2.4E-05 3.1E-04 3.1E-05 4.5E-05 2.5E-04 7.2E-04

1996 2.8E-04 4.2E-04 1.9E-05 1.7E-05 3.5E-04 1.3E-05 6.9E-05 9.8E-05 2.5E-04 8.2E-04

1997 6.8E-05 2.7E-05 1.3E-06 1.3E-05 0.0E+00 1.7E-06 6.0E-06 8.1E-05 6.1E-04

1998 3.6E-05 2.2E-05 1.7E-05 9.2E-05 2.9E-04

1999 1.7E-04 3.3E-04 1.2E-03

2000 0.0E+00 5.0E-04 5.9E-04 5.0E-04 9.7E-04 1.1E-03 4.2E-03

2001 0.0E+00 6.7E-04 1.9E-04 9.7E-04 6.7E-04 2.3E-04 4.2E-04 8.4E-04 4.1E-03

1 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 21 of 51

Uncertainty data presented in NTS annual environmental reports show that, in most instances, the 

standard deviation was less than 50% of the measured value. However, in a few cases, the standard 

deviation was as high as 100%.

For purposes of dose reconstruction, the GSD of the values in Table 4.2.1.2.2-2 can be estimated by 

using equation 4-1 (see Section 4.2.1.2.1) for a log-normal distribution. Under the favorable to 

claimant assumption that the 50th-percentile expected intakes are those in Table 4.2.1.2.2-2 and that 

the 95th-percentile values would be enveloped by a factor of ±2 (i.e., ±100%), then the GSD for the 

values in Tables 4.2.1.2.2-2 and 4.2.1.2.2-4 would be 1.52 for all values.

For purposes of reconstructing potential unmonitored dose, the annual intakes in Tables 4.2.1.2.2-2 

and 4.2.1.2.2-4 should be used only for individuals whose employment records indicate that they 

spent considerable time outdoors. Therefore, these intakes do not apply to miners or tunnel workers. 

In addition, reasonable adjustments to these values should be made whenever information indicates 

that the covered employee did not spend 100% of his workday in the outdoor environment. These 

individuals may have administrative job categories such as typist, secretary, janitor, or procedurewriter.

In addition, dose reconstructors should examine the values in Tables 4.2.1.2.2-2 and 4.2.1.2.2-4 and 

compare them with the values for resuspension intakes in Tables 4.2.2-2 and 4.2.2-3 for 238Pu and 

239, 240Pu because the tables in this section represent actual measured data, whereas the tables in 

Section 4.2.2 show calculated values based on favorable to claimant assumptions. Because both 

sets of tables represent the effects of the same primary pathway (i.e., resuspension), it would not be 

appropriate to use both sets of data to reconstruct a claimant dose. Although there is fairly good 

agreement between the two data sets, dose reconstructors should use the higher value for a given 

area when assessing intakes for unmonitored employees when the covered employee's work location

is well documented. If it is not known where a covered employee worked, the site average annual 

intakes for 238Pu and 239, 240Pu provided in Section 4.2.2, Table 4.2.2-3 should be used.

It is assumed that plutonium could be either S or M absorption type.

4.2.1.2.3 Gross Alpha

As discussed in Section 4.2.1.2, measurement of gross alpha in NTS atmosphere started in 1964 at 

12 locations. However, after a 5-day waiting period for decay of naturally occurring radon-thoron 

daughter products, long-lived alpha activity was infrequently detected. For this reason, and because 

of the high degree of uncertainty associated with alpha results, statistical summaries of the data were 

considered meaningless and therefore not reported in the annual environmental reports until 1971. In 

addition, the slight increases identified during these early years correlated well with fallout from 

foreign weapon tests and were therefore not attributed to NTS activities.

In 1970, REECo stated that plutonium results for air samples were uniformly low (i.e., on the order of 

1 pCi m-3). However, in the same report, REECo identified the area adjacent to the 9-300 crater 

(located in Area 9) as a source of slightly elevated, and rising, plutonium atmospheric concentration 

due to known plutonium fields that resulted from safety tests in the area before 1960.

Isotopic analysis of plutonium began in 1970 and, although data were provided for 239Pu, the annual 

environmental reports did not provide gross alpha data from 1974 to 1996. Because of the correlation 

of positive alpha activity with foreign weapon tests and the high statistical uncertainty associated with 

the early data, dose reconstructors should not use the data in Tables 4.2.1.2.3-1 and 4.2.1.2.3-2 to 

adjust environmental dose. These data are provided only for completeness.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 22 of 51

Table 4.2.1.2.3-1. Gross alpha atmospheric concentrations (pCi m-3) for sampled Areas. 

Area

Year 1 2 3 4 5 6 7 9 10 11 12 13

1971 3.3E-01 3.7E-01 3.3E-01 4.5E-01 3.5E-01 2.3E-01 2.9E-01 2.7E-01

1972 2.3E-01 2.0E-01 2.0E-01 1.8E-01 4.4E-01 3.7E-01 1.6E-01 1.8E-01 2.4E-01

1973 3.9E-02 4.1E-02 3.9E-02 3.5E-02 4.1E-02 1.2E-01 9.6E-02 3.7E-02 4.2E-02

1996 3.1E-03 4.0E-03 4.4E-03 3.3E-03 3.4E-03 3.2E-03 5.0E-03 6.5E-03 3.2E-03 3.4E-03 2.8E-03 4.8E-03

1997 1.5E-03 1.6E-03 1.9E-03 1.6E-03 1.8E-03 1.7E-03 1.5E-03 2.2E-03 1.6E-03 1.4E-03 1.1E-03 2.0E-03

1998 1.9E-03 1.6E-03 1.8E-03 1.5E-03 2.0E-03 1.8E-01 1.6E-03 1.9E-03 1.7E-03 1.5E-03 1.6E-03

1999 2.4E-03 2.3E-03 2.5E-03 2.4E-03 2.3E-03 2.2E-03 2.0E-03 4.6E-03 2.4E-03 2.1E-03

2000 1.5E-02 1.4E-02 2.6E-01 1.5E-02 1.5E-02 1.7E-02 1.6E-02 1.2E-01 1.5E-02

2001 1.9E-02 1.4E-02 2.4E-02 1.6E-02 1.4E-02 1.9E-02 1.6E-02 1.4E-02 1.4E-02

Table 4.2.1.2.3-1 (cont.). Gross alpha atmospheric concentrations (pCi m-3) for sampled Areas.

Area

Site 

average

Site 

maximum

Year 15 16 18 19 20 23 25 27 28

1971 2.9E-01 2.9E-01 3.1E-01 2.7E-01 4.1E-01 2.9E-01 4.5E-01 2.3E-01 4.5E-01

1972 6.7E-01 1.8E-01 3.8E-01 2.8E+00 1.9E-01 2.3E-01 2.7E-01 3.3E-01 2.8E+00

1973 3.4E-02 5.3E-02 1.3E-01 4.0E-02 4.5E-02 3.0E-02 4.7E-02 4.1E-02 1.3E-01

1996 5.4E-03 4.3E-03 3.4E-03 3.0E-03 3.2E-03 4.2E-03 3.2E-03 3.5E-03 6.5E-03

1997 1.5E-03 1.4E-03 1.5E-03 1.5E-03 1.6E-03 1.7E-03 1.3E-03 1.4E-03 2.2E-03

1998 1.6E-03 1.8E-03 1.7E-03 1.8E-03 1.6E-03 9.6E-03 1.8E-01

1999 2.5E-03 2.1E-03 2.1E-03 2.3E-03 1.6E-03 4.6E-03

2000 1.5E-02 1.7E-02 1.8E-02 1.5E-02 2.6E-02 2.6E-01

2001 1.8E-02 1.5E-02 1.6E-02 1.3E-02 1.6E-02 1.5E-02 1.2E-02 2.4E-02
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 23 of 51

Table 4.2.1.2.3-2. Gross alpha annual intakes (Bq) for sampled Areas. 

Year

Area

1 2 3 4 5 6 7 9 10 11 12 13

1971 3.0E+01 3.3E+01 2.9E+01 4.0E+01 3.1E+01 2.0E+01 2.5E+01 2.4E+01

1972 2.0E+01 1.8E+01 1.8E+01 1.6E+01 3.9E+01 3.3E+01 1.4E+01 1.6E+01 2.2E+01

1973 3.5E+00 3.6E+00 3.5E+00 3.1E+00 3.6E+00 1.1E+01 8.5E+00 3.3E+00 3.7E+00

1996 2.8E-01 3.6E-01 3.9E-01 2.9E-01 3.0E-01 2.8E-01 4.4E-01 5.8E-01 2.8E-01 3.0E-01 2.5E-01 4.3E-01

1997 1.3E-01 1.4E-01 1.7E-01 1.4E-01 1.6E-01 1.5E-01 1.3E-01 2.0E-01 1.4E-01 1.2E-01 9.8E-02 1.8E-01

1998 1.7E-01 1.4E-01 1.6E-01 1.3E-01 1.8E-01 1.6E+01 1.4E-01 1.7E-01 1.5E-01 1.3E-01 1.4E-01

1999 2.1E-01 2.0E-01 2.2E-01 2.1E-01 2.0E-01 2.0E-01 1.8E-01 4.1E-01 2.1E-01 1.9E-01

2000 1.4E+00 1.2E+00 2.3E+01 1.3E+00 1.4E+00 1.5E+00 1.4E+00 1.0E+01 1.4E+00

2001 1.7E+00 1.3E+00 2.2E+00 1.4E+00 1.2E+00 1.7E+00 1.4E+00 1.3E+00 1.2E+00

Table 4.2.1.2.3-2 (cont.). Gross alpha annual intakes (Bq) for sampled Areas.

Year

Area Site 

average

Site 

15 16 18 19 20 23 25 27 28 maximum

1971 2.5E+01 2.6E+01 2.7E+01 2.4E+01 3.6E+01 2.5E+01 4.0E+01 2.1E+01 4.0E+01

1972 5.9E+01 1.6E+01 3.4E+01 2.5E+02 1.7E+01 2.0E+01 2.4E+01 2.9E+01 2.5E+02

1973 3.0E+00 4.7E+00 1.1E+01 3.6E+00 4.0E+00 2.7E+00 4.2E+00 3.7E+00 1.1E+01

1996 4.8E-01 3.8E-01 3.0E-01 2.6E-01 2.8E-01 3.7E-01 2.8E-01 3.1E-01 5.8E-01

1997 1.3E-01 1.2E-01 1.3E-01 1.3E-01 1.4E-01 1.5E-01 1.2E-01 1.3E-01 2.0E-01

1998 1.4E-01 1.6E-01 1.5E-01 1.6E-01 1.4E-01 8.6E-01 1.6E+01

1999 2.2E-01 1.9E-01 1.9E-01 2.0E-01 1.4E-01 4.1E-01

2000 1.3E+00 1.5E+00 1.6E+00 1.4E+00 2.3E+00 2.3E+01

2001 1.6E+00 1.3E+00 1.4E+00 1.1E+00 1.4E+00 1.4E+00 1.0E+00 2.2E+00
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 24 of 51

4.2.1.2.4 Gross Beta

As discussed in Section 4.2.1.2, measurement of gross beta in the NTS atmosphere started in 1964 

at 12 locations. The gross beta concentrations in Table 4.2.1.2.4-1 are typically the average of the 

maximum concentrations for a given area in a given year. In cases where maximum values were not 

provided, the average of the average concentrations was reported. In addition to the annual area 

concentrations, Table 4.2.1.2.4-1 provides site average and maximum concentrations, which 

represent the arithmetic averages and maximums, respectively, of the concentrations of all site areas. 

Potential intakes associated with these concentrations can be calculated under the assumption that 

an unmonitored worker was occupationally exposed for 2,000 hr yr-1 with a breathing rate of 2,400 m3

yr-1. Table 4.2.1.2.4-2 presents these intakes.

For purposes of dose reconstruction, the GSD of the values in Table 4.2.1.2.4-2 can be estimated 

using equation 4-1 for a log-normal distribution. Uncertainty data in NTS annual environmental 

reports show that, in most instances, the standard deviation was less than 50% of the measured 

value. However, in a few cases, the standard deviation was as high as 100%.

Under the favorable to claimant assumption that the 50th-percentile expected intakes are those in 

Table 4.2.1.2.3-2 and that the 95th-percentile values would be enveloped by a factor of ±2 (i.e., 

±100%), then the GSD for the values in Tables 4.2.1.2.4-2 and 4.2.1.2.4-4 is 1.52 for all values.

For purposes of reconstructing potential unmonitored dose, the annual intakes in Table 4.2.1.2.4-2 

should be used only for individuals whose employment records indicate that they spent considerable 

time outdoors. Therefore, these intakes do not apply to miners or tunnel workers. 

In addition, dose reconstructors should examine the values in Table 4.2.1.2.4-2 and compare them 

with the values for resuspension intakes in Tables 4.2.2-2 and 4.2.2-3 for 90Sr (the unidentified 

radionuclide with the highest internal dose impact [i.e., the highest CEDE of all NTS beta emitters]). 

However, gross beta intakes in Table 4.2.1.2.4-2 during the early years (1967 through 1970) should 

not be used for adjusting internal dose because of the well-documented correlation of these elevated 

data with foreign weapons testing. 

Beginning in the early 1970s, the average and maximum annual intakes for 90Sr based on the 

measured data in Table 4.2.1.2.4-2 show good agreement with the calculated values for 90Sr in 

Table 4.2.2-3. However, because the tables in this section show actual measured data, whereas the 

tables in Section 4.2.2 show calculated values based on favorable to claimant assumptions (with the 

exception of the data before 1970) the annual intakes in Table 4.2.1.2.4-2 should be used to adjust 

internal dose to 90Sr for unmonitored employees. 

4.2.2

The resuspension of radionuclides after initial deposition on the ground surface is an exposure 

pathway that could, in certain situations, represent an important pathway to employees at NTS. 

Historical measurements at NTS and other locations have shown that the rate of resuspension 

decreases very rapidly with time, and that for conditions immediately following a test, resuspension is 

only of importance (compared to the inhalation exposure from the initial plume passage) over short 

periods. For cleanup and property-release situations, resuspension could be the dominant pathway, 

especially for transuranic radionuclides or for other radionuclides that do not readily cross biological 

barriers. Historically, the concern about resuspension has been with isotopes of plutonium, which do 

not readily cross biological boundaries but stay in the lung for long periods if inhaled (Anspaugh 2002, 

p. 675).

Annual Intakes from Resuspension
Effective Date: 04/20/2004 Revision No. 00

Document No. ORAUT-TKBS-0008-4 Page 25 of 51

Table 4.2.1.2.4-1. Gross beta atmospheric concentrations (pCi m-3) for sampled Areas. 

Area

Year 1 2 3 4 5 6 7 9 10 11 12

1966 2.1E+00 3.2E+00 1.7E+00 2.2E+00

1967 2.3E-01 3.4E+00 3.7E+00 2.2E+00 3.3E+00 3.0E+00 3.1E+00 2.5E+00 2.6E+00

1968 5.5E+01 4.7E-01 5.1E-01 7.5E-01 1.9E+00 7.1E+00 1.2E+00 6.3E-01 1.2E+00

1969 4.1E+01 1.1E+01 2.1E+01 4.0E+00 1.0E+00 1.3E+01 3.7E+00 1.2E+01

1970 8.7E-01 6.7E-01 7.1E-01 8.7E-01 8.4E-01 7.5E-01 7.8E-01 8.3E-01

1971

1972

1973

1974 1.3E+01 1.1E-02 9.7E-02 9.7E-02 1.1E-01 1.2E-01 9.6E-02 9.6E-02 1.3E-01

1975 9.1E-02 8.3E-02 9.4E-02 1.1E-01 1.0E-01 1.0E-01 8.5E-02 8.9E-02 1.1E-01

1976 9.3E-02 3.0E-02 3.8E-02 6.3E-02 4.9E-02 5.7E-01 1.8E-01 9.5E-02 4.0E-02

1977 1.7E-01 1.7E+00 1.6E-01 2.0E-01 2.6E-01 1.6E-01 1.2E-01 1.6E-01 1.6E-01

1978 1.1E-01 6.8E-02 8.9E-02 1.0E+01 9.8E-02 1.0E-01 9.7E-02 1.0E-01

1979 3.7E-02 3.5E-02 3.3E-02 3.6E-02 3.5E-02 3.1E-02 3.4E-02 3.4E-05 3.3E-02

1980 3.6E-02 4.0E-02 4.4E-02 3.8E-02 3.7E-02 2.4E-02 3.8E-02 3.9E-02 3.4E-02

1981 1.5E-01 1.6E-01 1.6E-01 1.6E-01 1.6E-01 1.7E-01 1.6E-01 1.7E-01 1.6E-01

1982 2.2E-02 2.2E-02 2.3E-02 2.4E-02 2.3E-02 2.4E-02 2.3E-02 1.7E-02 2.4E-02

1983 1.8E-02 1.8E-02 1.8E-02 1.9E-02 1.8E-02 1.7E-02 1.8E-02 1.9E-02 1.6E-02

1984 1.8E-02 1.8E-02 1.9E-02 2.2E-02 1.9E-02 1.8E-02 1.8E-02 1.7E-02 1.7E-02

1985 1.7E-02 1.6E-02 1.7E-02 2.1E-02 1.6E-02 1.6E-02 1.6E-02 1.7E-02 1.5E-02

1986 4.6E-02 5.2E-02 4.8E-02 5.6E-02 4.6E-02 4.7E-02 5.1E-02 4.8E-02 4.9E-02

1987 3.6E-02 3.9E-02 2.8E-02 2.1E-02 2.6E-02 2.4E-02 5.3E-02 1.9E-02 2.4E-02

1988 2.0E-02 2.6E-02 1.9E-02 2.2E-02 1.8E-02 1.8E-02 1.8E-02 1.9E-02 1.9E-02

1989 3.8E-02 4.5E-02 3.6E-02 1.9E-01 3.2E-02 5.8E-01 6.7E-02 3.3E-02 8.7E-02

1990 3.4E-02 3.2E-02 4.1E-02 4.2E-02 5.3E-02 4.5E-02 3.7E-02 3.3E-02 4.0E-02

1991 3.4E-02 7.7E-02 3.4E-02 4.2E-02 3.5E-02 3.4E-02 4.0E-02 3.3E-02 3.2E-02 4.5E-02

1992 3.8E-02 7.8E-02 4.6E-02 4.2E-02 3.7E-02 4.0E-02 4.6E-02 3.8E-02 3.8E-02 2.2E-01

1993 5.5E-02 3.5E-02 3.7E-02 4.3E-02 4.7E-02 3.3E-02 3.6E-02 3.4E-02 3.5E-02 3.2E-02

1994

1995 4.0E-02 4.3E-02 3.8E-02 4.0E-02 3.8E-02 4.5E-02 5.1E-02 4.6E-02 3.7E-02

1996 3.3E-02 3.5E-02 3.6E-02 3.5E-02 3.4E-02 3.4E-02 5.3E-02 2.9E-02

1997 1.9E-02 2.0E-02 1.9E-02 1.9E-02 2.2E-02 1.4E-02 1.8E-02 1.9E-02 2.0E-02 1.8E-02 1.6E-02

1998 2.0E-02 1.7E-02 2.0E-02 2.0E-02 1.8E-02 1.9E-02 1.6E-02 1.8E-02 1.8E-02

1999 2.2E-02 2.2E-02 2.1E-02 2.3E-02 2.1E-02 2.2E-02 2.1E-02 2.1E-02 2.2E-02

2000 5.1E-02 5.5E-02 5.4E-02 4.8E-02 3.7E-02 4.7E-02 4.0E-02 5.7E-02

2001 4.1E-02 4.1E-02 4.2E-02 2.7E-02 4.2E-02 3.9E-02 4.3E-02 3.9E-02 3.9E-02
Effective Date: 04/20/2004 Revision No. 00

Document No. ORAUT-TKBS-0008-4 Page 26 of 51

Table 4.2.1.2.4-1 (cont.). Gross beta atmospheric concentrations (pCi m-3) for sampled Areas.

Area Site 

average

Site 

Year 15 16 18 19 20 23 25 27 28 maximum

1966 2.2E+00 2.0E+00 1.5E+00 7.4E-01 3.2E+00

1967 7.8E+00 9.1E-01 4.8E+00 2.7E+00 2.6E+00 2.6E+01 3.4E+00 2.6E+01

1968 1.3E+00 8.9E-01 2.1E+00 3.5E+02 1.4E+00 6.9E-01 1.8E+00 2.1E+01 3.5E+02

1969 2.1E+01 1.6E+01 1.1E+00 1.1E+00 1.6E+00 1.8E+00 7.5E+00 4.1E+01

1970 7.5E-01 7.3E-01 7.6E-01 7.3E-01 1.1E+00 8.0E-01 9.5E-01 6.1E-01 1.1E+00

1971

1972

1973

1974 1.6E-01 7.8E-02 4.9E-02 8.3E-01 2.0E-01 9.9E-02 1.3E-01 7.5E-01 1.3E+01

1975 9.2E-02 9.4E-02 9.1E-02 1.3E-01 9.2E-02 9.3E-02 1.2E-01 7.9E-02 1.3E-01

1976 9.9E-02 4.9E-02 3.2E-02 1.0E-01 5.1E-02 2.8E-02 3.4E-02 7.8E-02 5.7E-01

1977 1.9E-01 1.7E-01 2.0E-01 5.2E-02 1.7E-01 1.4E-01 1.9E-01 2.0E-01 1.7E+00

1978 9.3E-02 9.2E-02 1.7E-01 8.9E-02 1.0E-01 1.1E-01 5.9E-01 1.0E+01

1979 2.4E-02 3.3E-02 3.2E-02 5.9E-02 3.5E-02 3.3E-02 3.5E-02 2.6E-02 5.9E-02

1980 3.8E-02 6.0E-02 6.6E-02 3.4E-02 3.8E-02 3.7E-02 3.0E-02 6.6E-02

1981 1.6E-01 1.5E-01 1.5E-01 3.9E-02 1.7E-01 1.7E-01 1.7E-01 1.2E-01 1.7E-01

1982 2.3E-02 2.2E-02 2.0E-02 2.0E-02 1.6E-01 2.2E-02 2.2E-02 2.2E-02 2.6E-02 1.6E-01

1983 1.7E-02 1.7E-02 1.6E-02 1.6E-02 2.3E-02 1.8E-02 1.7E-02 1.4E-02 2.3E-02

1984 1.5E-02 1.6E-02 1.6E-02 1.8E-02 1.8E-02 1.2E-02 2.2E-02

1985 1.5E-02 1.5E-02 1.4E-02 1.4E-02 1.8E-02 1.6E-02 1.2E-02 2.1E-02

1986 4.8E-02 3.6E-02 5.0E-02 4.8E-02 4.7E-02 4.8E-02 4.8E-02 3.8E-02 5.6E-02

1987 2.7E-02 1.6E-02 3.6E-02 2.8E-02 2.0E-02 1.4E-02 1.6E-02 2.1E-02 5.3E-02

1988 1.9E-02 1.7E-02 1.9E-02 2.0E-02 1.9E-02 1.7E-02 1.4E-02 2.6E-02

1989 2.5E-01 1.1E-01 1.2E-01 1.4E-01 4.6E-02 3.6E-02 3.4E-02 9.2E-02 5.8E-01

1990 3.6E-02 3.4E-02 3.3E-02 3.2E-02 3.9E-02 3.7E-02 4.6E-02 3.1E-02 5.3E-02

1991 3.2E-02 3.2E-02 3.1E-02 3.2E-02 4.4E-02 3.8E-02 3.7E-02 3.3E-02 7.7E-02

1992 1.5E-01 3.4E-02 3.4E-02 3.6E-02 3.6E-02 3.9E-02 3.8E-02 4.9E-02 2.2E-01

1993 3.7E-02 3.8E-02 3.3E-02 3.2E-02 1.9E-02 3.7E-02 3.9E-02 3.1E-02 5.5E-02

1994

1995 4.2E-02 3.7E-02 4.5E-02 2.8E-02 3.8E-02 4.0E-02 4.8E-02 3.7E-02 3.5E-02 5.1E-02

1996 3.4E-02 3.2E-02 3.3E-02 2.8E-02 3.5E-02 3.7E-02 3.3E-02 2.6E-02 5.3E-02

1997 1.8E-02 1.8E-02 1.8E-02 1.8E-02 2.7E-02 1.9E-02 1.7E-02 1.7E-02 2.7E-02

1998 1.9E-02 1.9E-02 2.0E-02 1.7E-02 1.2E-02 2.0E-02

1999 2.1E-02 2.0E-02 2.0E-02 1.9E-02 2.1E-02 1.5E-02 2.3E-02

2000 4.7E-02 7.0E-02 4.1E-02 4.8E-02 3.0E-02 7.0E-02

2001 3.5E-02 4.0E-02 3.7E-02 3.9E-02 5.4E-02 3.9E-02 3.0E-02 5.4E-02
Effective Date: 04/20/2004 Revision No. 00

Document No. ORAUT-TKBS-0008-4 Page 27 of 51

Table 4.2.1.2.4-2. Gross beta annual intakes (Bq) for sampled Areas. 

Area

Year 1 2 3 4 5 6 7 9 10 11 12

1966 1.8E+02 2.8E+02 1.5E+02 2.0E+02

1967 2.0E+01 3.0E+02 3.3E+02 2.0E+02 2.9E+02 2.7E+02 2.7E+02 2.2E+02 2.3E+02

1968 4.9E+03 4.2E+01 4.5E+01 6.6E+01 1.7E+02 6.3E+02 1.1E+02 5.6E+01 1.0E+02

1969 3.6E+03 1.0E+03 1.9E+03 3.5E+02 9.1E+01 1.1E+03 3.3E+02 1.1E+03

1970 7.7E+01 6.0E+01 6.3E+01 7.8E+01 7.5E+01 6.6E+01 6.9E+01 7.3E+01

1971

1972

1973

1974 1.1E+03 9.7E-01 8.6E+00 8.6E+00 9.7E+00 1.1E+01 8.5E+00 8.5E+00 1.1E+01

1975 8.0E+00 7.4E+00 8.4E+00 1.0E+01 8.9E+00 8.9E+00 7.6E+00 7.9E+00 9.8E+00

1976 8.3E+00 2.7E+00 3.4E+00 5.6E+00 4.4E+00 5.1E+01 1.6E+01 8.4E+00 3.6E+00

1977 1.5E+01 1.5E+02 1.4E+01 1.8E+01 2.3E+01 1.4E+01 1.1E+01 1.5E+01 1.4E+01

1978 9.6E+00 6.0E+00 7.9E+00 9.2E+02 8.7E+00 9.1E+00 8.6E+00 9.0E+00

1979 3.3E+00 3.1E+00 2.9E+00 3.2E+00 3.1E+00 2.8E+00 3.0E+00 3.0E-03 2.9E+00

1980 3.2E+00 3.6E+00 3.9E+00 3.4E+00 3.3E+00 2.1E+00 3.4E+00 3.5E+00 3.0E+00

1981 1.3E+01 1.4E+01 1.4E+01 1.4E+01 1.5E+01 1.5E+01 1.4E+01 1.5E+01 1.4E+01

1982 2.0E+00 2.0E+00 2.0E+00 2.2E+00 2.0E+00 2.1E+00 2.0E+00 1.5E+01 2.1E+00

1983 1.6E+00 1.6E+00 1.6E+00 1.6E+00 1.6E+00 1.5E+00 1.6E+00 1.7E+00 1.4E+00

1984 1.6E+00 1.6E+00 1.7E+00 2.0E+00 1.7E+00 1.6E+00 1.6E+00 1.5E+00 1.5E+00

1985 1.5E+00 1.4E+00 1.5E+00 1.9E+00 1.4E+00 1.4E+00 1.4E+00 1.5E+00 1.3E+00

1986 4.0E+00 4.6E+00 4.2E+00 4.9E+00 4.1E+00 4.2E+00 4.5E+00 4.3E+00 4.4E+00

1987 3.2E+00 3.5E+00 2.5E+00 1.9E+00 2.3E+00 2.1E+00 4.7E+00 1.7E+00 2.1E+00

1988 1.7E+00 2.3E+00 1.7E+00 2.0E+00 1.6E+00 1.6E+00 1.6E+00 1.7E+00 1.7E+00

1989 3.4E+00 4.0E+00 3.2E+00 1.7E+01 2.8E+00 5.2E+01 6.0E+00 2.9E+00 7.7E+00

1990 3.0E+05 2.8E+05 3.7E+05 3.7E+05 4.7E+05 4.0E+05 3.3E+05 2.9E+05 3.6E+05

1991 3.0E+00 6.8E+00 3.1E+00 3.7E+00 3.1E+00 3.0E+00 3.6E+00 2.9E+00 2.8E+00 4.0E+00

1992 3.4E+00 6.9E+00 4.1E+00 3.7E+00 3.3E+00 3.6E+00 4.1E+00 3.4E+00 3.4E+00 2.0E+01

1993 4.8E+00 3.1E+00 3.3E+00 3.8E+00 4.2E+00 2.9E+00 3.2E+00 3.0E+00 3.1E+00 2.8E+00

1994

1995 3.5E+00 3.8E+00 3.3E+00 3.6E+00 3.4E+00 4.0E+00 4.5E+00 4.0E+00 3.3E+00

1996 3.0E+00 3.1E+00 3.2E+00 3.1E+00 3.0E+00 3.0E+00 4.7E+00 2.6E+00

1997 1.7E+00 1.7E+00 1.7E+00 1.7E+00 1.9E+00 1.3E+00 1.6E+00 1.7E+00 1.7E+00 1.6E+00 1.4E+00

1998 1.8E+00 1.5E+00 1.8E+00 1.8E+00 1.6E+00 1.7E+00 1.4E+00 1.6E+00 1.6E+00

1999 2.0E+00 2.0E+00 1.8E+00 2.0E+00 1.9E+00 2.0E+00 1.9E+00 1.9E+00 2.0E+00

2000 4.5E+00 4.9E+00 4.8E+00 4.2E+00 3.3E+00 4.2E+00 3.6E+00 5.0E+00

2001 3.6E+00 3.6E+00 3.8E+00 2.4E+00 3.8E+00 3.5E+00 3.8E+00 3.5E+00 3.5E+00
Effective Date: 04/20/2004 Revision No. 00

Document No. ORAUT-TKBS-0008-4 Page 28 of 51

Table 4.2.1.2.4-2 (cont.). Gross beta annual intakes (Bq) for sampled Areas.

Area Site 

average

Site 

Year 15 16 18 19 20 23 25 27 28 maximum

1966 2.0E+02 1.7E+02 1.3E+02 6.6E+01 2.8E+02

1967 7.0E+02 8.0E+01 4.2E+02 2.4E+02 2.3E+02 2.3E+03 3.0E+02 2.3E+03

1968 1.1E+02 7.9E+01 1.9E+02 3.1E+04 1.3E+02 6.1E+01 1.6E+02 1.9E+03 3.1E+04

1969 1.9E+03 1.4E+03 9.9E+01 9.9E+01 1.4E+02 1.6E+02 6.7E+02 3.6E+03

1970 6.7E+01 6.5E+01 6.8E+01 6.5E+01 9.7E+01 7.1E+01 8.5E+01 5.4E+01 9.7E+01

1971 5.4E+01a 9.7E+01 a

1972 5.4E+01a 9.7E+01a

1973 5.4E+01a 9.7E+01a

1974 1.4E+01 6.9E+00 4.4E+00 7.4E+01 1.7E+01 8.8E+00 1.2E+01 6.7E+01 1.1E+03

1975 8.2E+00 8.4E+00 8.1E+00 1.2E+01 8.2E+00 8.3E+00 1.1E+01 7.0E+00 1.2E+01

1976 8.8E+00 4.4E+00 2.8E+00 9.1E+00 4.5E+00 2.5E+00 3.0E+00 6.9E+00 5.1E+01

1977 1.7E+01 1.5E+01 1.8E+11 4.6E+00 1.5E+01 1.3E+01 1.7E+01 1.8E+01 1.5E+02

1978 8.3E+00 8.1E+00 1.5E+01 7.9E+00 9.1E+00 9.5E+00 5.2E+01 9.2E+02

1979 2.1E+00 2.9E+00 2.8E+00 5.3E+00 3.1E+00 2.9E+00 3.1E+00 2.3E+00 5.3E+00

1980 3.4E+00 5.3E+00 5.9E+00 3.0E+00 3.4E+00 3.3E+00 2.7E+00 5.9E+00

1981 1.4E+01 1.3E+01 1.3E+01 3.4E+00 1.5E+01 1.5E+01 1.5E+01 1.1E+01 1.5E+01

1982 2.0E+00 2.0E+00 1.8E+00 1.8E+00 1.5E+01 2.0E+00 2.0E+00 2.0E+00 2.3E+00 1.5E+01

1983 1.5E+00 1.5E+00 1.4E+00 1.4E+00 2.0E+00 1.6E+00 1.5E+00 1.3E+00 2.0E+00

1984 1.4E+00 1.4E+00 1.4E+00 1.6E+00 1.6E+00 1.1E+00 2.0E+00

1985 1.4E+00 1.3E+00 1.2E+00 1.2E+00 1.6E+00 1.4E+00 1.1E+00 1.9E+00

1986 4.3E+00 3.2E+00 4.4E+00 4.3E+00 4.2E+00 4.2E+00 4.3E+00 3.4E+00 4.9E+00

1987 2.4E+00 1.4E+00 3.2E+00 2.5E+00 1.8E+00 1.2E+00 1.4E+00 1.9E+00 4.7E+00

1988 1.6E+00 1.5E+00 1.7E+00 1.8E+00 1.7E+00 1.5E+00 1.3E+00 2.3E+00

1989 2.2E+01 9.8E+00 1.1E+01 1.2E+01 4.1E+00 3.2E+00 3.0E+00 8.2E+00 5.2E+01

1990 3.2E+05 3.0E+05 2.9E+05 2.8E+05 3.5E+05 3.3E+05 4.1E+05 2.7E+00 4.7E+00

1991 2.8E+00 2.8E+00 2.7E+00 2.8E+00 3.9E+00 3.3E+00 3.3E+00 2.9E+00 6.8E+00

1992 1.3E+01 3.0E+00 3.0E+00 3.2E+00 3.2E+00 3.5E+00 3.4E+00 4.4E+00 2.0E+01

1993 3.3E+00 3.4E+00 2.9E+00 2.8E+00 1.7E+00 3.3E+00 3.5E+00 2.8E+00 4.8E+00

1994 2.8E+00b 4.8E+00 b

1995 3.7E+00 3.3E+00 4.0E+00 2.5E+00 3.4E+00 3.6E+00 4.3E+00 3.3E+00 3.1E+00 4.5E+00

1996 3.0E+00 2.8E+00 2.9E+00 2.5E+00 3.1E+00 3.2E+00 2.9E+00 2.3E+00 4.7E+00

1997 1.6E+00 1.6E+00 1.6E+00 1.6E+00 2.4E+00 1.6E+00 1.5E+00 1.5E+00 2.4E+00

1998 1.7E+00 1.6E+00 1.8E+00 1.5E+00 1.1E+00 1.8E+00

1999 1.9E+00 1.8E+00 1.8E+00 1.7E+00 1.9E+00 1.3E+00 2.0E+00

2000 4.2E+00 6.2E+00 3.7E+00 4.3E+00 2.6E+00 6.2E+00

2001 3.1E+00 3.6E+00 3.3E+00 3.4E+00 4.8E+00 3.4E+00 3.1E+00 3.6E+00 2.7E+00 4.8E+00

a. Value not provided in annual report - assumed same as for 1970.

b. Value not provided - assumed same as for 1993.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 29 of 51

Airborne concentrations of radionuclides resuspended from contaminated soils can be estimated 

based on the following expression (Till and Meyer 1983, p. 5-30):

CA = RF × CS eq. (4-2)

where

CA = air concentration, Bq m-3 

CS = radionuclide areal soil deposition, Bq m-2 

RF = resuspension factor, m-1 

Extensive studies were performed in the 1980s to quantify residual contamination at NTS (McArthur 

1983, 1985, 1987, 1988, 1989, 1991). Table 2-6 of the Site Description (Section 2) for this Technical 

Basis document lists the results of these studies. Table 4.2.2-1 lists the total areal depositions based 

on the inventory values in Table 2-6 (see Section 2 of this document) divided by the respective areal 

size. 

Historically, there have been several models used to estimate the resuspension of radionuclides from 

contaminated soil. However, the most widely used model is the resuspension-factor model shown by 

equation 4-2. Efforts to estimate the resuspension factor, RF, at NTS began as early as 1957 and 

have culminated with a predictive model by Anspaugh (2002). This model (shown below as equation 

4-3) was derived from an expanded databank of evaluated measurements (Beck 1980; Hicks 1982) 

and has attempted to describe empirical measurements over the entire span of the data set (i.e., 

early, intermediate, and late periods). The model also includes an explicit statement of uncertainty, a 

factor of 10, which is large but is indicative of the dispersion in the data sets. The model is as follows:

t0705 t00309 9 11m10x10e10x6e10SF −− −− − −± = [ + + ] ).( ).( eq. (4-3)

Figure 4.2.2-1 shows the time-dependent nature of the resuspension factor as a function of time after 

initial deposition. As shown, after about 3 years the resuspension factor approaches a value of about 

1.3 × 10-9 m-1. Because the vast majority of radionuclides in surface soils were deposited as a result 

on atmospheric testing that stopped in 1962, and because it is reasonable to assume that 

unmonitored employees would not be likely to be exposed to freshly deposited radionuclides, the 

value of 1.3 × 10-9 m-1 is assumed for estimating potential airborne concentrations to which employees 

could have been exposed. 

Application of this resuspension factor with areal soil deposition values in Table 4.2.2-1 allows for 

estimation of airborne radionuclide concentrations at various locations at NTS. If the assumption is 

made that the standard breathing rate of an employee is 2,400 m3 yr-1, these concentrations can be 

used to estimate annual intakes. Table 4.2.2-2 lists these annual intakes.

For situations in which a covered employee spent a large amount of the working day outside, but the 

exact locations are not known, dose reconstructions should use the site average annual intake values 

in Table 4.2.2-3. 

For purposes of dose reconstruction, the GSD of the values in Table 4.2.2-2 can be estimated using 

equation 4-1. Under the assumption that the 50th-percentile expected intakes are those in Tables 

4.2.2-2 and 4.2.2-3 and that the 95th-percentile values would be enveloped by a factor of ±10, then 

the GSD for the values in Tables 4.2.2-2 and 4.2.2-3 are 4.1 for all values. 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 30 of 51

Table 4.2.2-1. Radionuclide areal soil deposition (Bq m-2). 

Area

Area 

(m2) Am-241 Pu-238

Pu239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 6.9E+07 2.3E+03 3.5E+03 1.3E+04 5.9E+02 4.7E+03 8.1E+03 8.1E+03 5.4E+01 2.7E+02

2 5.1E+07 2.1E+03 6.2E+03 1.6E+04 8.7E+02 1.7E+04 3.3E+04 1.0E+04 2.9E+02

3 8.4E+07 2.0E+03 1.4E+03 1.6E+04 4.4E+02 5.3E+03 1.5E+04 8.0E+03 4.4E+01 2.2E+02

4 4.1E+07 5.9E+03 1.2E+04 3.6E+04 1.4E+03 1.1E+04 1.2E+04 8.1E+03 1.8E+02

5 7.5E+06 3.0E+03 4.9E+02 2.4E+04 3.0E+03 2.0E+03 4.4E+03 4.9E+04 9.9E+02

6 8.4E+07 7.5E+02 1.5E+03 3.7E+03 8.8E+01 1.2E+03 1.5E+03

7 5.0E+07 1.6E+03 4.4E+02 1.2E+04 7.4E+02 3.8E+03 6.8E+03 1.6E+04 1.5E+02 2.2E+02

8 3.6E+07 1.7E+04 8.2E+03 1.1E+05 5.9E+03 4.3E+04 2.6E+04 4.5E+03 6.2E+02

9 5.2E+07 3.0E+03 1.6E+03 6.4E+04 5.0E+02 6.2E+03 9.3E+03 1.6E+04 1.4E+02 2.1E+02

10 5.2E+07 1.4E+04 1.4E+04 7.9E+04 6.9E+03 6.0E+04 3.9E+04 1.6E+03 2.1E+02 3.6E+03

11 1.0E+07 1.2E+04 1.8E+03 1.0E+05 0.0E+00 1.8E+03 1.1E+03

12 1.0E+08 2.1E+03 3.1E+03 1.4E+04 4.3E+02 7.2E+03 6.1E+03

15 9.1E+07 3.2E+03 3.2E+03 2.5E+04 1.2E+02 7.7E+03 8.9E+03

16 3.7E+07 7.0E+02 1.5E+03 3.7E+03 1.0E+02 2.9E+03 3.7E+03

17 8.1E+07 1.3E+03 2.0E+03 8.2E+03 4.5E+02 6.8E+03 8.6E+03

18 7.1E+07 9.9E+03 2.9E+03 5.2E+04 3.7E+02 5.2E+03 8.9E+03 5.8E+02 5.2E+01 4.2E+02

19 3.8E+08 2.0E+03 3.1E+03 1.3E+04 1.1E+02 3.5E+03 3.0E+03

20 1.6E+07 5.3E+04 6.9E+04 9.4E+04 1.8E+04 1.3E+04 9.9E+03 3.0E+04 3.7E+03 1.1E+04

25 2.3E+06 3.2E+03 1.6E+03 6.3E+03

26 5.2E+05

30 7.8E+05 1.5E+05 2.1E+05 6.7E+05 3.8E+04 7.1E+04 6.2E+04 3.3E+04 4.8E+03 9.5E+03

Resuspension Factor

1.3 E-09 m-1

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

0 100 200 300 400 500 600 700 800 900 1000

Days

meter-1

Figure 4.2.2-1. Resuspension factor as a function of time after initial deposition. 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 31 of 51

Table 4.2.2-2. Average annual intakesa (Bq) for NTS Areas from soil resuspension. 

Area Am-241 Pu-238

Pu239,240 Co-60 Cs 137 Sr-90 Eu-152 Eu-154 Eu-155

1 7.06E-03 1.09E-02 4.04E-02 1.85E-03 1.48E-02 2.52E-02 2.52E-02 1.68E-04 8.41E-04

2 6.56E-03 1.95E-02 4.98E-02 2.72E-03 5.43E-02 1.04E-01 3.17E-02 9.05E-04

3 6.35E-03 4.28E-03 5.11E-02 1.38E-03 1.66E-02 4.55E-02 2.48E-02 1.38E-04 6.90E-04

4 1.84E-02 3.62E-02 1.11E-01 4.46E-03 3.34E-02 3.62E-02 2.54E-02 5.57E-04

5 9.22E-03 1.54E-03 7.38E-02 9.22E-03 6.15E-03 1.38E-02 1.54E-01 3.07E-03

6 2.35E-03 4.55E-03 1.16E-02 2.76E-04 3.86E-03 4.83E-03

7 5.08E-03 1.39E-03 3.70E-02 2.31E-03 1.20E-02 2.12E-02 5.08E-02 4.62E-04 6.93E-04

8 5.45E-02 2.57E-02 3.53E-01 1.83E-02 1.35E-01 8.02E-02 1.41E-02 0.00E+00 1.92E-03

9 9.36E-03 4.90E-03 1.98E-01 1.56E-03 1.94E-02 2.90E-02 5.13E-02 4.46E-04 6.69E-04

10 4.23E-02 4.23E-02 2.45E-01 2.16E-02 1.87E-01 1.23E-01 4.90E-03 6.69E-04 1.11E-02

11 3.68E-02 5.57E-03 3.23E-01 0.00E+00 5.57E-03 3.34E-03

12 6.42E-03 9.57E-03 4.39E-02 1.35E-03 2.25E-02 1.91E-02

15 1.01E-02 9.85E-03 7.95E-02 3.79E-04 2.40E-02 2.78E-02

16 2.18E-03 4.68E-03 1.15E-02 3.12E-04 9.04E-03 1.15E-02

17 3.97E-03 6.39E-03 2.56E-02 1.42E-03 2.13E-02 2.70E-02

18 3.10E-02 9.14E-03 1.63E-01 1.14E-03 1.63E-02 2.78E-02 1.80E-03 1.63E-04 1.31E-03

19 6.31E-03 9.62E-03 4.21E-02 3.31E-04 1.08E-02 9.32E-03

20 1.65E-01 2.16E-01 2.95E-01 5.68E-02 3.95E-02 3.09E-02 9.35E-02 1.15E-02 3.45E-02

25 9.91E-03 4.95E-03 1.98E-02

26

30 4.75E-01 6.69E-01 2.08E+00 1.19E-01 2.23E-01 1.93E-01 1.04E-01 1.49E-02 2.97E-02

a. Geometric standard deviation is assumed to be 4.1 for all values.

Table 4.2.2-3. Site average and 

maximum annual intakesa (Bq) for NTS 

from soil resuspension. 

Radionuclide

Average

intake

Maximum

intake,

Am-241 1.31E-02 4.75E+00

Pu-238 1.39E-02 6.69E+00

Pu-239 7.93E-02 2.08E+01

Co-60 3.13E-03 1.19E+00

Cs-137 2.71E-02 2.23E+00

Sr-90 2.87E-02 1.93E+00

Eu-152 1.16E-02 1.54E+00

Eu-154 2.53E-04 1.49E-01

Eu-155 1.19E-03 3.45E-01

a. Geometric standard deviation is assumed to be 

4.1 for all values.

For purposes of reconstructing potential unmonitored dose, the annual intakes in Tables 4.2.2-2 and 

4.2.2-3 should be used only for individuals whose employment records indicate that they spent 

considerable time outdoors. Therefore, these intakes do not apply to miners or tunnel workers. In 

addition, reasonable adjustments to these values should be made whenever information indicates the 

covered employee did not spend 100% of his workday in the outdoor environment.

In addition, dose reconstructors should examine the values derived from measured data in Tables 

4.2.1.2.2-2 and 4.2.1.2.2-4 (Section 4.2.2.2) and compare them with the values for resuspension 

intakes in Tables 4.2.2-2 and 4.2.2-3 for 238Pu and 239, 240Pu because the tables in this Section 4.2.1.2 

represent actual measured data, whereas the tables in this section show calculated values based on 

favorable to claimant assumptions. Because both sets of tables represent the effects of the same 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 32 of 51

primary pathway (i.e., resuspension), it would not be appropriate to use both sets of data to 

reconstruct a claimant dose. Although there is fairly good agreement between the two data sets, dose 

reconstructors should use the higher value for a given area when assessing intakes for unmonitored 

employees when the covered employee's work location is well documented. If it is not known where a 

covered employee worked, the site average annual intakes for 238Pu and 239, 240Pu provided in Section 

4.2.2, Table 4.2.2-3 should be used to assess annual intakes for each year of employment. 

With the exception of plutonium, the vast majority of the radionuclides in NTS soil resulted from 

atomic weapons tests and were thereby exposed to extreme heat. For purposes of calculating organ 

dose using the Integrated Modules for Bioassay Analysis program, internal dose reconstructors 

should assume biological clearance factors are Type S. Because much of the plutonium released 

was a result of safety tests, these radionuclides should not be considered high-fired. Therefore, for 

plutonium, dose reconstructors should use maximizing absorption factors for either Type S or Type M.

4.2.3

There are several major factors that could affect the accuracy of the average annual intake estimates 

in Tables 4.2.2-2 and 4.2.2-3. The most important factors include the following:

Uncertainty

• Uncertainty in and time-dependent nature of the resuspension factor

• Spatial variations in radionuclide soil concentrations

• Duration of exposures

Each of these factors could result in and over- or underestimation of annual intakes equal to several 

orders of magnitude. The following sections discuss each of these factors.

4.2.3.1 Time-Dependent Nature of the Resuspension Factor

As discussed in Anspaugh (2002), the uncertainty suggested in equation 4-3 is plus or minus one 

order of magnitude. This is a large uncertainty, but it is based on actual measurements that suggest 

the very complex nature and unpredictability of the resuspension process. In addition, the 

time-dependence of the resuspension factor suggested by equation 4-3 covers a range of almost four 

orders of magnitude. Therefore, if an unmonitored employee was exposed to freshly deposited 

fallout, the intake values in Tables 4.2.2-2 and 4.2.2-3 could represent significant underestimates of

the actual intake. However, because of the rapid decrease in the resuspension factor with time, the 

likelihood of exposure of this type for periods greater than a few days would be very small. In 

addition, a favorable to claimant factor of 10 has been applied to the resuspension factor to minimize 

the likelihood that airborne concentrations would be underestimated. In addition, the number of 

unmonitored employees likely to have been inadvertently exposed to fresh fallout would be small.

4.2.3.2 Spatial Variations in Radionuclide Soil Concentrations

The values of radionuclide areal soil concentration in Table 4.2.2-1 are average values for each 

particular area. The actual sample values for locations within each of the areas can vary by several 

orders of magnitude. In addition, McArthur (1991, p. 24) estimated the uncertainty of the inventory 

measurements at a factor of ±2. Therefore, the intake values estimated from these average soil 

concentrations could under- or overestimate the actual intake values depending on the location of the 

unmonitored employee within a particular area. However, it is reasonable to assume that radiological 

control practices would have prevented unmonitored employees from being inadvertently exposed in 

the more highly contaminated locations.
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4.2.3.3 Duration of Exposures

The intake values in Tables 4.2.2-2 and 4.2.2-3 are based on the assumption that the unmonitored 

employee was exposed for an occupational year (i.e., 2,000 hr). This assumption, although favorable 

to claimants, is probably an overestimate of the actual time an unmonitored employee would likely 

have spent outdoors during a year. In addition, this assumption does not take into consideration any 

protection from airborne dust particles that could have been afforded by vehicles during transportation 

to and around work locations. Taken together, these factors could result in an overestimation of 

annual intake by a factor of 10 or more.

4.3 OCCUPATIONAL EXTERNAL DOSE

Workers incur external doses from the ambient radiation levels and from noble gases. Ambient 

radiation measurements were reported in NTS annual environmental reports (see reference section 

for document citations) starting in 1967; however, no ambient radiation data were provided in the 

annual environmental reports between 1968 and 1976. 

4.3.1

Before 1967, ambient radiation levels that were unaffected by weapons testing were not reported in 

the annual environmental reports. Although there were many radiation measurements completed 

between 1951 and 1967, most of these were to characterize the effects of weapons tests and were 

therefore not appropriate for use in estimating external environmental dose for unmonitored 

employees.

Ambient Radiation

In 1967, ambient radiation levels were measured using Victoreen Model 239 indirect reading 

ionization chambers with effective ranges of 0 to 10 mR hr-1. Five of these chambers were in small, 

semiprotective enclosures at locations chosen in NTS living areas. 

As a backup for the ionization chambers, standard NTS film dosimeters were included at each sample 

location. However, all results from these film dosimeters for 1967 were either zero or lost due to light 

or heat damage. Results from environmental NTS film dosimeters were subsequently discarded 

because of the likelihood of heat damage.

The ionization chambers were collected on a weekly basis and read on a Victoreen Minometer II 

reader. Corrections were made for background and for nonradiation induced drift. Two sets of 

chambers were used, one set at the sample locations for measurements and another set stored fully 

charged in the laboratory. Each week these sets were exchanged, the fresh set being recharged and 

a record kept of the amount of "drift" while stored in the laboratory. A specially designed shock-proof 

box for transportation minimized accidental discharge from mechanical shock.

Readings from the five chambers in each location were averaged to obtain a mean value for each 

location each week. Readings significantly higher than others at a particular location were not used in

compiling the data because the abnormal readings were most likely the result of shock or other 

malfunction and were not representative measurements. Because of the method of background 

subtraction discussed above, the data in the 1967 annual report were assumed to be due to 

manmade radiation in excess of natural background.

No results were reported between 1968 and 1976 in the NTS annual environmental reports, however, 

the ambient radiation reporting was reestablished in 1977 using thermoluminescent dosimeters 

(TLDs). The dosimeters used were CaF2:Dy (TLD-200) 0.25- by 0.25- by 0.035-in. chips from 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 34 of 51

Harshaw Chemical Company. A badge consisting of at least two chips shielded by 0.047 in. of 

cadmium (1030 mg cm-2) inside a 0.050-in. black plastic (140 mg cm-2) holder was placed about 1 m 

above the ground at each of 10 locations that coincided with air sampling stations. These sites were 

selected because of their proximity to workers. During that year, the natural background ambient 

radiation level was established at 0.26 mrem d-1 or about 95 mrem yr-1. 

In 1987, the Harshaw dosimeters were replaced with Panasonic Model UD-814 TLDs. These 

specifically designed environmental dosimeters contain three identical CaSO4:Tm elements and one 

Li2B4O7:Cu element. The lithium element is shielded with 14 mg cm-2 of material to monitor beta 

particles in the environment. The three calcium elements are encapsulated in 1,000 mg cm-2 of 

plastic and lead to monitor ambient gamma levels. The UD-814 TLDs have remained in use to 

present time.

In subsequent years, the number of sampling locations increased to over 150 and eventually covered 

all populated areas at NTS. Table 4.3.1-1 lists average ambient background radiation levels. Table 

4.3.1-2 lists the annual ambient radiation measured with background included and site average and 

maximum values. For purposes of dose reconstruction, because the values in Table 4.3.1-2 

represent continuous exposure for an entire year, these values should be adjusted for occupational 

exposure (i.e., 2,000 hr yr-1), and added to the dose of record for unmonitored employees. If the area 

in which the employee worked is known, the average value for that area should be used. If the area 

in which the employee worked is not known, the favorable to claimant site maximum value should be 

used.

It should also be noted that some covered employees remained onsite continuously for weeks at a 

time. However, because most of the non-working hours were spent indoors where elevated ambient 

radiation would be reduced by shielding from the building, and because of the conservative 

assumptions used to estimate the values in Table 4.3.1-2, adjustment of the tabular data is not 

required to assure ambient exposures are not underestimated for these individuals.

It is likely that the elevated ambient radiation peaked in the early 1960s near the end of above-ground 

testing. Therefore, it would be reasonable and probably favorable to claimants to assign maximum 

annual environmental doses for the years 1963 through 1966 equal to the maximum value reported 

for 1967 (i.e., 318 mrem yr-1). It should be noted that for 1967, the maximum value is about 50 

percent greater than the average value for all areas where measurements were reported. Following 

this logic, maximum doses have been added to Table 4.3.1-2 (shown in italics) for 1969 through 1977 

for which data were not readily available. These values were based on the maximum value for 1978 

and should be used to adjust environmental dose for unmonitored employees for those years.

Although not explicitly stated in NTS annual environmental reports, the overall uncertainty of 

measurements made with modern TLDs (e.g., Panasonic UD-814) has been determined to be less 

than 20%. However, for earlier measurement methods, the overall uncertainty could have been 

greater. Therefore, for purposes of dose reconstruction, the favorable to claimant assumption is 

made that the 95th-percentile values of the expected values, 50th percentile, in Tables 4.3.1-1 and 

4.3.1-2 are enveloped by a factor of ±2 of the overall error of 50% (i.e., ±100%) Then equation 4-1 

(Section 4.2.1.2.1) would estimate the GSD for the expected values at 1.52.
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 35 of 51

Table 4.3.1-1. Ambient 

background radiations 

levels by year. 

Year mrem yr-1

1964 (a)

1965 (a)

1966 (a)

1967 (a)

1968 (a)

1969 (a)

1970 (a)

1971 (a)

1972 (a)

1973 (a)

1974 (a)

1975 (a)

1976 (a)

1977 95

1978 95

1979 95

1980 99

1981 110

1982 100

1983 100

1984 100

1985 100

1986 72

1987 102

1988 131

1989 106

1990 110

1991 112

1992 109

1993 131

1994 93

1995 94

1996 91

1997 95

1998 88

1999 91

2000 115

2001 114

a. Not reported.
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Table 4.3.1-2. Ambient radiation (mrem yr-1) by Area.

a

Area

Year123456789 10 11 12 13 14

1967 318 194 205

1968 285 183 175 190

1969

1970

1971

1972

1973

1974

1975

1976

1977

1978 130 167 200 138 110 110 120 150 320 140 150

1979 119 147 190 140 133 104 125 115 140 168 155 191

1980 123 156 199 155 133 106 130 135 145 192 165 196

1981 120 155 218 150 166 113125 150 180 160 187

1982 126 149 188 145 141 108 125 140 209 140 189

1983 115 141 181 143 144 103 95 135 201 140 174

1984 97 149 162 116 111 86 327 125 102 166 112 155

1985 141 78 101 107 347 94 114 225 180

1986 93 126 150 92 92 77 31894 96 120 107 106

1987 146 189 184 164 191 126 120 149 181 133 168

1988 155 209 207 164 177 134 150 179 203 158 194

1989 140 217 205 132 141 106 126 151 179 153 139

1990 134 167 187 142 142 109 122 148 170 132 140

1991 128 168 194 146 165 116 132 124 132 150 130 190

1992 123 167 173 141 150 104 145 128 136 157 158 170

1993 147 178 200 158 198 134 165 153 156 172 184 201

1994 62 62 89 84 70 102 97

1995 102 136 150 121 212 87 176 102 104 179 124 130

1996 102 131 151 110225 105 164 98 99 176 124 135

1997 131 86 88 81

1998 106 132 137 110 134 100 201 127 101 168 121 148

1999 108 133 135 122 110 105 191 129 99 160 122 152

2000 127 155 161 129 134 122 217 147 190 187 144 176

2001 121 299 151 120 136 107 214 143 115 180 140 163
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Table 4.3.1-2 (cont.). Ambient radiation (mrem yr-1) by Area.

Area

Year 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 Average Maximum

1967 142 175 207 318

1968 186 131 172 189 285

1969 320b

1970 320b

1971 320b

1972 320b

1973 320b

1974 320b

1975 320b

1976 320b

1977 320b

1978 129 173 185 150 162 140 120 155 320

1979 121 150 145 155 160 115 117 152 130 120 125 140 191

1980 134 165 163 185 207 70 98 163 135 130 205 152 207

1981 138 155 168 174 163 75 126 175 150 140 152 218

1982 134 150 173 176 200 75 95 150135 135 140 146 209

1983 124 135 151 158 206 125 163 149 140 185 148 206

1984 108 129 124 132 157 58 154 129 173 137 327

1985 115 123 131 127 103 153 143 347

1986 81 124 134 160 55 71 106 89 135 115 318

1987 135 151 174 246 204 68 96 129 139 155 246

1988 155 172 212 223 211 84 123 163 172 223

1989 162 160 168 178 189 78 116 146 118 150 217

1990 124 164 174 178 192 83 177 139 146 148 192

1991 122 165 175 173 189 81 79 143 154 110 155 144 194

1992 134 152 168 189 185 79 75 132 143 179 145 189

1993 145 177 195 192 213 91 96 158 169 195 167 213

1994 143 141 131 58 61 115 121 140 98 143

1995 109 129 137 149 153 58 70 117 121 127 212

1996 103 137 144 153 150 55 56 114 124 165 128 225

1997 102 150 153 83 63 127 135 170 114 170

1998 99 135 151 156 76 58 113 123 165 127 201

1999 108 137 153 151 75 57 111 126 170 126 191

2000 117 160 185 182 92 81 139 158 205 153 217

2001 112 167 159 131 90 74 135 145 193 147 299

a. Blank cells indicate data not available.

b. Ambient radiation not provided for these years - assumed equal to value for 1978.
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4.3.2

Ground seepage can increase when changes in ambient pressure pump small amounts of noble 

gases (primarily 85Kr and 133Xe) up through the overburden and into the atmosphere from the cavity 

created by a nuclear test. This process, sometimes referred to as atmospheric pumping, creates a 

diffuse source of radiological effluents.

Releases of Noble Gases

In 1982, REECo assumed responsibility for six noble gas sampling stations previously run by EPA 

and replaced them with new samplers. These sampling units are housed in a metal tool box with 

three metal air bottles attached with quick disconnect hoses. A vacuum is maintained on the first 

bottle, which causes a steady flow of air to be collected in the other two bottles. The flow rate is 

approximately 0.5 cm3 min-1. The two collection bottles are exchanged weekly and yield a sample 

volume of about 3 × 105

 cm3. 

The noble gases are separated and collected from the atmospheric sample by a series of 

cryogenic-gas chromatographic techniques. Water and carbon dioxide are removed at room 

temperature and the krypton and xenon are collected on charcoal at liquid nitrogen temperatures. 

These gases are transferred to a molecular sieve where they are separated from any remaining gases 

and each other. The krypton and xenon are transferred to separate scintillation vials and counted on 

a liquid scintillation counter. The lower limits of detection for krypton and xenon are 4 × 10-6 and 

10 × 10-6 pCi m-3, respectively.

Krypton-85

The original six samplers were in Areas 1, 5, 12, 15, 23, and 25. These six samplers showed no 

statistical difference in atmospheric concentrations of 85Kr during 1982 and 1983. The average 

concentration of all stations for these years was 24.6 pCi m-3 (0.91 Bq m-3), which was established as 

NTS background. This background measurement shows good agreement with the global background 

of 27 pCi m-3 (1.0 Bq m-3) (WMO 2002).

In 1983, REECo measured a statistically significant elevation in 85Kr concentrations in Area 20 of 

4.0 pCi m-3 (0.15 Bq m-3) above background, which was subsequently related to seeps from nuclear 

cavities on Pahute Mesa. These elevated concentrations continued until after the cessation of 

nuclear testing in 1992. Table 4.3.1-3 lists the average background and the Area 20 elevated 

concentrations from 1983 through 1992.

The source of the elevated 85Kr concentrations was attributed to the emplacement hole nearest to the 

Area 20 sampler (U-20a), which is about 3,660 m (12,000 ft) to the south. To determine whether 

these elevated concentrations could be of any consequence for unmonitored workers, upper-bound 

concentrations of 85Kr within 500 m of U-20a were estimated using Area 20 meteorological data 

station files (Grossman 2003) and the CAP88-PC (EPA 1998) computer program. The program was 

used to develop atmospheric dispersion factors, X/Qs, for various distances from the source of 

emissions. The ration of the X/Q value at 500 m to that at 3,660 m was determined to be about 27, 

which means that the 85Kr concentration 500 m from the source could, on average, be 27 times 

greater than that measured at the sampler location. The potential dose resulting from submersion in 

the plume of this noble gas is considered to be entirely external because the gas is not readily 

assimilated by the body. Therefore, these source concentrations were converted to potential organ 

doses using submersion dose conversion factors from Federal Guidance Report No. 12 (EPA 1993). 

As shown in Table 4.3.1-3, the upper-bound annual organ doses for all years are less than 1 mrem. 

For purposes of reconstructing environmental doses for unmonitored employees, these doses are 

inconsequential for determination of probability of causation and therefore should be ignored. In 
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Table 4.3.1-3. 85Kr average background and the Area 20 net concentrations and annual organ doses 

from 1983 through 1992. 

Year

NTS 

background 

(pCi m-3)

Area 20 

Camp 

(pCi m-3)

Net 

elevated 

(pCi m-3)

Sourcea

elevated 

concentration 

(pCi m)

Sourcea

elevated 

concentration 

(Bq m-3)

Other 

organb

dosec at 

sourcea

(mrem yr-1)

Skin dosed

at sourcea

(mrem yr-1)

1983 27 46 19 502 18.6 0.0029 0.18

1984 25 29 4 106 3.9 0.0006 0.04

1985 28 39 11 290 10.8 0.0017 0.10

1986 30 58 28 739 27.4 0.0043 0.26

1987 26 39 13 343 12.7 0.0020 0.12

1988 25 29 4 106 3.9 0.0006 0.04

1989 23 27 4 106 3.9 0.0006 0.04

1990 29 37 8 211 7.8 0.0012 0.07

1991 24 32 8 211 7.8 0.0012 0.07

1992 26 30 4 106 3.9 0.0006 0.04

1993 27 28 1 26 1.0 0.0002 0.01

a. Source receptor location assumed to be 500 m from U-20a emplacement hole.

b. Includes gonads, breast, lung, red marrow, bone surfaces, thyroid, remainder, and effective whole body.

c. Assumes 2,000 hr yr-1 exposure and dose conversion factor of 2.2 × 10-16 Sv per Bq s m-3 (EPA 1993).

d. Assumes 2,000 hr yr-1 exposure and dose conversion factor of 1.32 × 10-14 Sv per Bq s m-3 (EPA 1993).

addition, if a monitored employee was exposed to the calculated upper-bound 85Kr concentrations or 

significantly higher concentrations during work activities, these exposures would have been monitored 

by personal dosimeters and would, therefore, already be present in the individual dosimetry record.

Uncertainty estimates provided for a number of the 85Kr measurements indicate, generally, a standard 

deviation of less than 20% of the measured value. Even at the 2 sigma level of ±40 percent, the 

annual organ doses in Table 4.3.1-3 are less than 1 mrem and therefore inconsequential for 

employee dose reconstruction.

Xenon-133

Although the vast majority of the 133Xe measurements were below the lower limit of detection (i.e., 

10 × 10-6 pCi m-3), elevated atmospheric concentrations of 133Xe were occasionally measured in Area 

20. Like 85Kr, these elevated concentrations were attributed to seepage from the underground test 

cavities at the Pahute and Rainier Mesa tests. However, 133Xe has a short half-life (5.25 d versus 

10.7 yr for 85Kr). Therefore, these elevated concentrations were short-lived. In addition, these 

elevated 133Xe concentrations were always less than the derived air concentrations (typically less than 

3%) and would result in annual organ doses of less than 1 mrem. Therefore, potential doses to 

unmonitored employees from elevated concentrations of 133Xe are inconsequential for dose 

reconstruction.

4.4 OCCUPATIONAL DOSE TO TUNNEL WORKERS FROM EXPOSURE TO RADON

In the early 1980s, the Environmental Sciences Department of REECo recognized that the buildup of 

radon and radon daughter concentrations (RDCs) could pose a potential health problem in tunnels on 

Rainer Mesa and at other locations at the Nevada Test Site. In 1984, to determine the concentrations 

of the RDCs and the effect of environmental conditions on the buildup of these concentrations, 

REECo conducted radon measurement surveys in G-, T-, and N-Tunnels (Favor 1987). This section 

discusses the results of these surveys and those conducted in 1991 and 1992 (Lyons 1992a,b).
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4.4.1

Area 12, which is in the Nuclear or High Explosive Test Zone, occupies 104 km2 (40 mi2

) at the 

northern boundary of NTS, known as Rainier Mesa. No atmospheric nuclear tests have occurred at 

this location; however, Area 12 was the site of the U.S. Atomic Energy Commission’s first fully 

contained underground nuclear detonation, named the Rainier test, on September 19, 1957, in a 

horizontal tunnel about 487 m (1,600 ft) into the mesa and 274 m (900 ft) beneath the top of the 

mesa. In the past several decades, a number of tunnels have been mined into Rainier Mesa, in which 

most of the U.S. Department of Defense’s (DOD’s) horizontal line-of-sight exposure experiments have 

occurred. The N-, P-, and T-tunnel complexes, in particular, were developed extensively during the 

1970s and 1980s. The tunnel experiments usually involved complex construction of large-diameter 

(up to 9 m [27 ft]), line-of-sight pipes and special closure mechanisms, blast and gas seal doors, 

stemming plugs, and the like. The G-tunnel complex was originally established for nuclear testing 

purposes but, since 1971, has been used only as an underground research facility. 

Tunnel Activities

In addition to its use for nuclear testing purposes, N-tunnel was the location of a Nonproliferation 

Experiment involving 1.3 million kg (2.9 million lb) of conventional explosives on September 22, 1993. 

DOD operates a high-explosives research and development tunnel in Area 12. This reusable test bed 

supports programs involving the detonation of conventional or prototype high explosives and 

munitions. 

The U1a Complex in Area 1 is an underground laboratory of horizontal tunnels about 0.5 mi in length 

at the base of a vertical shaft about 960 ft beneath the surface. The vertical shaft is equipped with a 

mechanical hoist for personnel and equipment access. Another vertical shaft about 1,000 ft away 

provides cross ventilation and instrumentation as well as utility access and emergency egress. The 

shaft was excavated in the 1960s, and a nuclear test was conducted in a horizontal tunnel mined from 

its base in 1990.

4.4.2

Radon-222, with a radioactive half-life of 3.8 d, occurs in the 238U decay chain; 220Ra, with a 

radioactive half-life of 54.5 s, occurs in the 232Th chain. The decay daughters associated with 222Rn 

include 218Po, 214Pb, 214Bi, and 214Po, and those associated with 220Ra include 216Po, 212Pb, 212Bi, and 

208Tl). Because radon is a noble gas and thus chemically inert, it migrates easily from the tunnel rock 

and soil, which contains naturally occurring trace quantities of uranium and thorium.

Radon Measurements

Radon daughter measurements in N- and T-Tunnels were made in drifts mined in a rock formation 

known as Tunnel Bed Non-Welded Ash Fall Tuff. This rock is unconsolidated and only slightly 

fractured. The drift sampled in the inclined G-Tunnel was mined in Grouse Canyon Welded Ash Fall 

Tuff, which is extremely fractured. Some factors that affect concentrations of radon and radon 

daughters in air are ventilation rates, barometric pressure, relative humidity, temperature inversions, 

the degree of fracturing in the rock, and the amount of smoke and dust in the air. Factors such as 

barometric pressure and fractures affect the rate at which radon emanates from the rock, while others 

such as dust and ventilation rates affect the accumulation rate of daughter products in the air. 

Concentrations fluctuate with changing seasons, changing weather conditions, and changing activities 

in the area being monitored.

The concentration of radon daughters (the major dose contributors) in air is measured in working 

levels (WL). This is the common unit for expressing radon decay product exposure rates. The WL 

was developed for use in uranium mines but is now also used for environmental exposures. 

Numerically, the WL is any combination of short-lived decay products in 1 L of air that will result in the 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 41 of 51

emission of 1.3 × 105 MeV of potential alpha energy. When radon is in complete equilibrium with its 

short-lived decay products, 1 WL equals 100 pCi L-1 (i.e., 100 pCi L-1 each of 222Rn and short-lived 

decay products 218Po, 214Pb, 214Bi, 214Po) (NCRP 1988, p. 17). For 220

Rn and its decay daughters of 

216Po, 212Pb, 212Bi, and 208Tl, 1 WL is equal to 7.47 pCi L-1. The advantage of the WL unit is that it 

allows comparison of different equilibrium levels and different concentrations of radon decay products. 

The degree of equilibrium is a critical factor for estimating inhalation exposure and is of equal 

importance to the radon concentration itself (NCRP 1998, p. 19). The WL unit considers this factor. 

However, because the dose conversion factors for 220Rn daughters (Pb-212 + Po-212) are one-third 

the values for the daughters of 222Rn (ICRP 1981), the dose equivalent WL for 220Rn is only one-third 

that of a 222

Rn WL.

The exposure of tunnel workers can be expressed in units of working-level months (WLM), which is 

an exposure rate of 1 WL for a working month of 170 hr (NCRP 1988, p. 17). For example, an 

exposure of 1 WLM would result from exposure to a concentration of 1 WL for 1 month or 0.5 WL for 

2 months. 

4.4.3

Measurements were taken in tunnels N, T, and G. The preliminary measurements indicated N- and 

T-Tunnel RDCs of about 0.01 WL under normal ventilation conditions. However, the data 

demonstrated that RDCs can rise to relatively high levels (i.e., 0.24 WL) when ventilation rates were 

significantly lower (Favor 1987, Figure 2, p. 7). The radon daughter concentrations in G Tunnel were 

an order of magnitude higher than those in N and T Tunnels. The average RDC in the rock 

mechanics drift (the worst-case location in G Tunnel) was 0.13 WL (ranging from 0.07 to 0.23 WL. 

Elevated RDCs in the rock mechanics drift of G Tunnel seemed to be from a lower ventilation rate in 

conjunction with the more highly fractured nature of the welded tuff in which the incline drift was 

mined. By increasing the ventilation rate, a 60% reduction in RDCs from an average of 0.13 WL to an 

average of 0.05 WL was achieved (see Table 4.4.3-1). 

Tunnel Radon Concentrations

Lyons (1992a,b) reported additional radon measurements from 1991 and 1992. Although the 

G-Tunnel complex remained inactive during 1992, radon samples were taken to document potential 

radon WL in a worst-case scenario of complete ventilation failure throughout the complex. The 

maximum WL for radon daughters was 1.4. No samples were taken during 1992 in T-Tunnel, which 

was inactive that year.

Table 4.4.3-1. Results of experiment to determine the effects of 

ventilation conditions on RDCs in the G Tunnel inclined drift. 

Location

Ventilation 

conditions

Radon daughter grab 

sample average (WL)

Rock mechanics drift at 0+52 Alternating 0.13

Continuous 0.05

Rock mechanics access drift at 

the heating block alcove

Alternating 0.13

Continuous 0.04

Bottom of rock mechanics 

access drift

Alternating 0.13

Continuous 0.04

EV6/EV5 Junction Alternating 0.1

Continuous 0.03

EV5 at HF36 Alternating 0.08

Continuous 0.03

Average Alternating 0.13

Continuous 0.05

Source: Favor (1987, Table 3).
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 42 of 51

Tables 4.4.3-2 and 4.4.3-3 list the results of extensive radon sampling in N- and P-Tunnel complexes 

in 1992 (from Lyons 1992a, Table 3, p. 8; and Lyons 1992b, Table 1, p. 8, respectively). The average 

concentrations in N- and P-Tunnels were 0.021 and 0.009 WL, respectively, and the maximum 

concentrations in N- and P-Tunnel were 0.038 and 0.017 WL, respectively.

Table 4.4.3-2. Radon daughter concentrations for N-Tunnel in 1991 and 

1992. 

N Tunnel location

Rn-222 WL Rn-220 WL

Average Maximum Average Maximum

January–June 1992

Miner's Lunchroom 0.005 0.007 0.01 0.016

Raytheon Alcove 0.003 0.004 0.005 0.008

Slow Alcove 0.005 0.007 0.008 0.009

24 Bypass Drift 0.005 0.006 0.015 0.015

24 LOS Drift at GZ 0.005 0.007 0.012 0.014

22 Bypass 0.006 0.007 0.013 0.015

July–December 1991

Miner's Lunchroom 0.005 0.007 0.011 0.018

Raytheon Alcove 0.009 0.03 0.017 0.06

21 LOS at 2 + 50 0.034 0.059 0.029 0.046

15 Assembly Drift 0.006 0.009 0.015 0.021

Slow Alcove 0.004 0.008 0.01 0.025

23 Fast Alcove 0.009 0.014 0.018 0.03

24 Bypass Drift 0.005 0.007 0.012 0.041

24 LOS Drift 0.005 0.01 0.015 0.036

Average 0.008 0.013 0.014 0.025

Table 4.4.3-3. Radon daughter concentrations for P-Tunnel in 1991 and 

1992. 

P-Tunnel location Rn-222 WL Rn-220 WL

Average Maximum Average Maximum

January–June 1992

01 Drift at Access Drift 0.001 0.001 0.002 0.005

01 Fast Alcove 0.001 0.001 0.004 0.006

02 Main Drift at 6 + 00 0.001 0.001 0.002 0.004

04 Reentry 0.001 0.002 0.004 0.005

04 LOS at 12 + 00 0.002 0.002 0.005 0.006

HPD Base Station 0.001 0.001 0.003 0.006

Miner's Lunchroom 0.001 0.001 0.003 0.006

05 Cavity 0.001 0.001 0.003 0.004

July–December 1991

01 Drift at Access Drift 0.003 0.006 0.004 0.007

01 Fast Alcove 0.003 0.005 0.003 0.006

02 Main Drift at 6 + 00 0.01 0.032 0.01 0.046

04 LOS at VP X-Cut 0.008 0.013 0.012 0.015

04 LOS Drift at GZ 0.005 0.006 0.014 0.02

04 LOS Test Ch. 0.003 0.004 0.005 0.006

04 Bypass at RE#1 0.007 0.013 0.008 0.011

IHD Alcove 0.003 0.009 0.004 0.015

LLNL Alcove 0.003 0.006 0.007 0.02

05 Cavity 0.003 0.006 0.006 0.015

Average 0.003 0.006 0.006 0.011
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 43 of 51

4.4.4

Although measurements were periodically performed in the tunnel complexes to ensure adequate 

worker protection, neither the U.S. Department of Energy nor its predecessor agencies attempted to 

quantify or record occupational exposures to radon and its daughters. Therefore, dose reconstructors 

should adjust the dose of any employee who was a miner or tunnel worker to account for radon 

exposure while working in the tunnel complexes. To quantify the exposure, Table 4.4.4-1 lists 

favorable to claimant airborne 222

Rn RDCs based on measurement results. It should be noted that the 

values provided in Table 4.4.1-1 pertain only to 222

Rn and its daughter products. The 220Rn WL 

concentrations and resultant exposures for N- and P-Tunnels must be converted to annual alpha 

organ dose prior to entry into IREP.

Tunnel Workers Exposure to Radon

Table 4.4.1-1. Annual 222Rn exposures and uncertainties for internal dose 

reconstruction for miners and tunnel workers. 

Tunnel 

complex

RDC concentration 

(WL)

Annual exposure

(WLM)a

Uncertainty

GSDb

G (before 1984) 0.13 1.56 1.33

G (1984 and 

later) 0.05 0.60 1.33

N 0.013 0.16 1.33

P 0.006 0.07 1.33

T 0.01 0.12 1.33

U1a Complexc 0.05 0.60 1.33

Unidentifiedd 0.05 0.60 1.33

a. Based on 170 hr of exposure per month for 12 months.

b. Based on the assumption that one standard deviation equals ±30%.

c. Applicable to underground work only.

d. Use these values when underground work location is not known.

For Tunnel G workers, annual exposures (assuming a full year of underground activity) are assumed 

to be 1.56 WLM before 1984 and 0.60 WLM during and after 1984. This is because REECo did not 

recognize until 1984 that significant reductions in RDCs could be effected by leaving all ventilation 

fans running overnight. Until 1984, the practice was to shut down alternating fans each night, which 

resulted in higher average RDCs.

Because no radon measurement data were readily available for the underground portions of the 

U1a Complex, the favorable to claimant assumption was made that the RDCs in U1a would be similar 

to those in G Tunnel (which is mined in highly fractured tuff). In addition, a reasonable assumption 

was made that the U1a Complex is well ventilated, similar to G Tunnel after 1984. Dose 

reconstructors should use these average, well-ventilated, annual exposure values for miners and 

tunnel workers without identified tunnel locations.

For purposes of dose reconstruction, the GSD of the values in Table 4.4.4-1 can be estimated using 

equation 4-1. Under the assumption that the 50th-percentile expected annual exposures are those in 

Table 4.4.4-1 and that the 95th-percentile values would be enveloped by a factor equal to ±60 percent 

(i.e., two times a sigma value of ±30 percent), then the GSD for the values in Tables 4.4.4-1 are 1.33

for all values. 

4.4.5

Two methods used to measure the concentration of radon and its progeny were the grab sample 

technique and the integration technique. Preliminary measurements were made by grab sampling for 

Uncertainty
Effective Date: 12/08/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-4 Page 44 of 51

a general estimate of the concentrations. Lucas cells collected and counted grab samples of radon 

gas. The cells were evacuated in the laboratory and opened at the sampling location for a single 

intake of air. The filled cells were counted at the EPA Las Vegas Laboratory within 24 hr to minimize 

decay. The number of measurements was small due to limited availability of Lucas cells. Grab 

samples of radon daughters were collected and counted according to a technique described by Rolle 

(ANSI 1973), which utilized a single count of an air sample collected on a filter. Samples were 

collected on a 2-cm Whatman fiberglass filter at a rate of approximately 5 L min-1 and counted on an 

EDA Instruments Incorporated Radon Detector (Model #RD200). The Rolle method allowed for a 

choice of several different analysis regimes. The regime chosen for tunnel measurements was a 

5-min sampling time, a 6-min decay time, and a 5-min counting time. Using appropriate correction 

factors, the RDC in WL can be evaluated 11 min after sample collection. The values obtained using 

the Rolle method were periodically checked using the Kusnetz (Favor 1987, p.3) method, which also 

utilized a single count of an air filter but allowed for a 40- to 90-min decay time. These measurement 

techniques have been found to have good accuracy and a relative standard deviation of less than 

15% (Favor 1987, p. 3). 

Grab sampling is quick and convenient. However, as mentioned above, radon concentrations can 

fluctuate widely with time and location. To account for this variability, integrating monitoring 

instruments were used. A passive environmental radon monitor (PERM) was used to measure radon 

in pCi L-1. A radon progeny integrating sampling unit (RPISU) measured the RDC in WL.

The PERM is an integrating radon monitor that employs electrostatic collection of radon daughter ions 

and uses a TLD as the radiation dose integration element. The lower limit of detection at the 95% 

confidence level is approximately 0.3 pCi L-1 for a 1-week sample with a relative standard deviation of 

20% (Favor 1987, p. 4). The RPISU is a low-volume air sampler that draws air through a 0.65-μmpore-size Millipore membrane filter. A Teflon disc TLD containing dysprosium-activated calcium 

fluoride (Harshaw TLD2OO) is positioned close to the collection surface of the filter and serves as the 

dose integrating element. The lower limit of detection at the 95% confidence level is approximately 

0.0001 WL for a 1-week sampling period with a relative standard deviation of 10% (Favor 1987, p. 4). 

The RPISU technique is recommended for the measurement of RDCs by the Technical 

Measurements Center of the U.S. Department o f Energy, Division of Remedial Action Projects (Favor 

1987, p. 4).

4.5 INSTRUCTIONS TO THE DOSE RECONSTRUCTOR

4.5.1

Starting in April of 1957, all personnel entering the NTS were monitored for external radiation 

exposure. Because unexposed control badges were processed together with the personnel 

dosimeters and the readings from the control badges were subtracted from the dosimeter to obtain a 

net reading for determining exposure, ambient external dose should not be included in dose 

reconstruction after this time. Before the institution of universal badging in April 1957 for those 

employees who were not monitored, coworker doses should be assigned based on the values in 

ORAUT (2006a, Table 6-11), and no additional ambient dose should be assigned. Based on a review 

of the dosimetry literature of the site, no change in the handling of dosimeters occurred when 

universal badging was introduced; therefore, ambient dose is considered to be appropriately included 

in the coworker doses.

Assigning External Ambient Dose

External ambient doses for each year are available in ORAUT (2006b) and can be used for workers 

who have continuous employment but nonexistent or incomplete monitoring records. For example, 

EG&G employees were likely stationed in Las Vegas and received dosimeters when they traveled to 
Effective Date: 12/08/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-4 Page 45 of 51

NTS. If the employee’s location cannot be clearly identified, it might be appropriate to assign ambient

dose for intervals with no dosimeter exchanges. This would be favorable to the claimant, but the 

assignment of ambient dose is not a requirement. A second example includes REECo employees 

who were processed as new employees through NTS (Mercury) and assigned to work at TTR. These 

employees have a minimum of one NTS dosimeter. Often no additional records are provided for 

these employees by Sandia National Laboratory, the responsible organization for dosimetry at TTR. 

For these REECo employees assigned at TTR, ambient dose from NTS can be assigned as 

appropriate. 
Effective Date: 04/20/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-4 Page 46 of 51

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